ML18022A568

Semiannual Radioactive Effluent Release Rept,Jan-June 1987. W/870831 Ltr

ML18022A568
Person / Time
Site:	Harris
Issue date:	06/30/1987
From:	Mcduffie J, Poland A, Watson R CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
CON-NRC-579 NUDOCS 8709030467
Download: ML18022A568 (149)

Text

REGULATOf'NFORMATION DISTRIBUTION TEM (R IDS) g ACCESSION 8709'030467 DQC. DATE'7/06/30 NOTARIZED: NO DOCKET 0 IL: 50-400 NBR'AC Hart is Nuclear Power Planti Unit ii Carolina

'hearon 05000400

~AUTH. NAME AUTHOR AFFILIATION POLAND'. Q. Caro 1 ina P oeer Sc Lig h t Co.

/AC DUFF I E i J. W. Carolina Power Sc Light Co.

WATSONi R. A. Carolina Poeer 5 Light Co.

REC IP. NAME RECIPIENT AFFILIATION

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"Semiannual Radioactive Effluent Release Rept'an-June f'UBJECT:

19'87 " W/870831 ltr.

DISTRIBUTION CODE: IE48D COPIES RECEIVED: LTR ENCL SIZE:

TITLE: 50. 3&a(a) (2) Semiannual Effluent Release Reports NOTES: Application for permit reneeal filed. 05000400

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Carolina Power & Light Company HARRIS NUCLEAR PROJECT P. 0 ~ Box 165 New Hill, North Carolina 27562 AUG 5 3 l987 File Number'SHF/10-13510C Letter Number. HO-870490 (0)

U.S. Nuclear Regulatory Commission NRC-579 ATTN: NRC Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT 1 DOCKET NO. 50-400 LICENSE NO. NPF-63 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT Gentlemen.'n accordance with Technical Specification 6.9.1.4, the Seminannual Radioactive Effluent Release Report is attached for the Shearon Harris Nuclear Power Plant. This report covers the period from initial criticality (January 3, 1987) through June 30, 1987.

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Very truly yours, R. A. Watson Vice President Harris Nuclear Project ONH:skm Attachment cc'Messrs. Dr. J. Nelson Grace (NRC RII)

Mr. B. Buckley (NRR)

Mr. G. Maxwell (NRC SHNPP)

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Carolina Power h Light Shearon Harris Nuclear Power Plant License No. NPF.-063 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY 1, 1987 to JUNE 30, 1987 Prepared by:

Project Specialist Radiation Control Reviewed by:

Manager E ronme al h Radiation Control Approved by:

Plant General Manager

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Table of Contents Page No.

Introduction i Discussion Appendix 1. Supplemental Information

l. Regulatory Limits 1/1

2. MPC's and dose rates which determine 1/2 maximum instantaneous rates.

34 Methods for Approximations of 1/2 Total Radioactivity

4. Batch Releases 1/3

5. Unmonitored Releases 1/3 Appendix 2. Effluent and Waste Disposal Report

1. Lower Limits of Detectability (LLD's) 2/1

2. Effluents Released 2/3

3. Solid Waste Disposal 2/10 Appendix 3. Changes to Process Control Program 3/1 Appendix 4. Changes to Offsite Dose Calculation Manual Appendix 5. Changes to Environmental Monitoring Program

l. Environmental Monitoring Program 5/1

2. Land Use Census 5/2 Appendix 6. Additional Technical Specification Responsibilities

l. Inoperability of Liquid Effluent Monitors 6/1

2. Inoperability of Gaseous Effluent Monitors 6/3

3. Unprotected Outdoor Tanks Exceeding Limits 6/5

4. Gas Storage Tanks Exceeding Limits 6/6 Appendix 7. Major Modifications to Radwaste System 7/1

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Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Introduction This Semiannual Radioactive Effluent Release Report is submitted per Technical Specification 6;9.1.4 to the Shearon Harris Nuclear Power Project (SHNPP) Operating License No. NPF-63. This is the first semiannual release report submitted in fulfillment of the plants Radiological Effluent Technical Specification (RETS). This reporting requirement was effective beginning with initial criticality, which occured on January 3,,1987.

However, with one exception discussed under Appendix 6 of the following section, the data in this report actually commences on January 1, 1987. This was done for consistency with future reporting periods and because 'the RETS were fully implemented as of that date.

Discussion

~A pendices 1 and 2:

The information on gaseous and liquid effluents is given in accordance with Regulatory Guide 1.21 (Rev. 1) Appendix B format. No solid waste was shipped during this period so no data is reported.

Activity concentrations (uCi/ml) and total curies released are for only those nuclides that were positively identified. If no activity for a nuclide is reported for a quarter, the Lower Limit of Detection (LLD) table shows a typical sensitivity level for detection of the nuclide.

No activity above background was detected in any potential continuous liquid release pathway. Therefore the summations of liquid effluents are based entirely on nuclide analysis and volume determinations of batch releases.

These results are based on methodology in the Offsite Dose Calculation Manual (ODCM).

Gaseous effluent activities for Quarter 1 were estimated from results of nuclide analyses of monthly stack gas grab samples and stack flow rate estimates based on design fan flow rates. Problems with the stack flow monitor calibrations and the flow integrator system rendered most of the release rate (uCi/sec) data stored on the RM-21 report processor computer invalid. However, the gas grab sampling and flow rate estimating methods are in accordance with Tech Spec alternative actions and provided suitable estimates of effluent release quantities, especially since the plant was primarily in low power testing modes during this quarter.

Although the flow monitor problems persisted through most of Quarter 2, improved data collection of hourly average stack monitor readings (in uCi/ml) was started. This data combined with the stack flow estimates provided more continuous accountability of stack effluents.

The gross activity concentrations above background were apportioned into specific nuclide amounts using the relative amounts detected in successive gas grab sample analyses. This methodology, although cumbersome when done manually, as it was during the 2nd Quarter, is identical to the method the

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HM-21 computer would have used had the stack flow input to the system been valid. It should be noted that the cuties reported are considered to be significantly overestimated because of the use of design fan flow rates which have consistently been found to be higher than actual flows. For the 2nd Quarter, the use of conservatively low background monitor readings for determining the net activity released also contributed to overestimating the curies released.

Appendix 3:

No changes to the Process Control Program (PCP) were made during this report period.

AP2 endix 4:

Changes made to the ODCM during this report period are listed. All changes were reviewed and approved by the Plant Nuclear Safety Committee (PNSC).

These changes do not reduce the accuracy or reliability of the dose calculations or monitor setpoint determinations.

Appendix 5:

No changes were made to the Environmental Monitoring Program during this report period. Changes to the Land Use Census are given based on a May 1987 survey. New census data is provided for distances to nearest special locations and for meat animal types nearest to SHNPP.

Appendix 6:

All effluent monitor inopexabilities greater than 30 days are given along with a brief explanation. Per prior agreement with the NRC, similar inoperable monitor periods prior to initial criticality and after receipt of the Operating License (October 24, 1986) are also given. No unprotected outdoor tank or gas storage tank exceeded Tech Spec limits during this report period.

Appendix 7:

The changes made to the radwaste processing system are described. These changes received the required 10CFR50.59 safety review and will not result in any increased exposure to the general public. Revised quantities of radwaste expected to be generated compared to those given in the FSAR are provided.

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1. Regulatory Limits A. Fission and activation gases (1) Calendar Quarter

a. 5 mrad gamma

b. 10 mrad beta (2) Calendar Year

a. l0 mrad gamma

b. 20 mrad beta B. I-131, I-133, I-135, H-3 and particulates with half-lives greater than eight days (1) Calendar Quarter

a. 7.5 mrem to any organ (2) Calendar Year

a. 15 mrem to any organ C. Liquid effluents (1) Calendar Quarter

a. 1.5 mrem to total body

b. 5 mrem to any organ (2) Calendar Year

a. 3 mrem to total body

b. 10 mrem to any organ

tl I 4 4 i I Il "'t' I II lh",II It Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 1: Supplemental Information

2. Maximum permissible concentrations and dose rates which determine maximum instantaneous rates.

A. Fission and activation gases (1) 500 mrem/year to total body (2) 3000 mrem/year to the skin B. I-131, I-133, I-135, H-3 and particulates with half-lives greater than eight days. 1500 mrem/year to any organ C. Liquid effluents The concentration of radioactive material released in liquid effluents to unrestricted areas after dilution shall be limited to the concentration specified in 10CFR20, Appendix B, Table II, Column 2, for radionuclides other than noble gases. (1) Tritium: MPC = 3.0E-3 uCi/ml; and (2) Dissolved and Entrained gases: MPC = 2.0E-4 uCi/ml

3. Measurements and Approximations of Total Radioactivity A. Fission and activation gases Measurements by continuous monitors, analysis by gamma spectroscopy and liquid scintillation counting for specific radionuclides in representative grab samples times total stack flow.

B. Iodines Measurements by continuous monitors and analysis by gamma spectroscopy for specific radionuclides collected on charcoal cartridges times total stack flow. C. Particulates Measurements by continuous monitors, analysis by gamma spectroscopy, alpha counting and radiochemical analysis for specific radionuclides collected on filter papers times total stack flow. D. Liquid Effluents Analysis by gamma spectroscopy and liquid scintillation counting for specific radionuclides by. individual releases. 1/2 l)r<<<<*>; ,) 7)r>F )v lr rr,erg(<<'; 1 I)' g'>> l<<>[II >>1 )l ." I*' '>l (( <<f le)>> ! [M<") )),)">>>> ~ ((<<I[ ~ .'I[i> ) ~ <<) ) ll ~ I tfl)) f I) ' ) "),' 1)')[ f,<<" ) '>>!P >Pf[ 0 r,m(>> .,F[1 Ff l) (( I 'r>I I f t)IF 'F.M( f'7,[' I l<< >[I ((lilt M (i"l(l I I ', / ) > l'[IF II l )<< 'r.l> )I l "i h )F;I ' <<)I"(I "I,, I, I ((( ~ <<"..(<F '[tf . I,)Ill ~ I '"'" II I 1 [( ([>>>>[M/r ( f " <<f r 'l I [Fr<< h I . '>> [lii, )g I . ', '[r l'1',>1<<., h)rr<< <<r, I th >>F<<l, >, = l, (>>(((1 ' <<h>> " <<('I ".Mf~ I' I> I 'gr> r I') 'FF 1 ( >> "I IFI(<<,', <<MM(lg)I(t (I7;,i ( 1 )l ( h, I" '>> "" 1" tj.'" ) "<<tl>>. ( ' ' W () k<<r'",' *'1) ) ' ll,(g,'h(gh " i ) f." ( " ""I !l >> ([... )'> ll'>> 1 ' II <<1 f<<) ~> )FFMI=t([" I )I((M ) .I ')' II)>' '!,(') ref )1 tf ~ (([lll>>" I ) ['>> (hh>> '>>'(I" .'." <<<<'"('I(I l(" , 1 1 I"F)f),',' it, h fil<< ' I << ) n <<(;F1, > 1 ~ it l<<1 l <<M ',I', r ~ I[> W <<r> Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 1: Supplemental Information

4. Batch Releases A. Liquid (1) Number of batch releases: 5.02 E+02 (2) Total time period for batch releases: 1.56 E+05 min.

(3) Maximum time for a batch release: 8.97 E+02 min. {4) Average time for a batch release: 3.12 E+02 min. (5) Minimum time for a batch release: 1.00 E+00 min. {6) Average stream flow during periods of release: 2.12 E+03 gpm B. Gaseous {1) Number of batch releases: 1.40 E+Ol (2) Total time period for batch releases.'3) 3.18 E+03 min Maximum time for a batch release: 1.25 E+03 min. (4) Average time for a batch release: 2.27 E+02 min. (5) Minimum time for a batch release: 2.30 E+Ol min.

5. Abnormal Releases A. Liquid No abnormal liquid releases were made in the period.

B. Gaseous No abnormal gaseous releases were made in the period. ,$ 'f I a na II t a ),Ia(;a I,, ~a " Va I')a,',V'ff ~ fl ,a aa V aa 11 ff, g )f ffif *V V/ j)g) g -av vl V ff )') ') ~ Vi f . <<) ) ')1 ) a ~ VV)fal <<) v 'f f .,g () la,,'VV 1'Ij V V <<lla)'>ra iIV" a) "l fi V')pl V iiCI a"'t aa I ~ g)'ll If, 1llf aaa 1V ' f "g ~ if ff )g) fl

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l. LLD's for Gaseous Effluents NUCLIDE LLD (uCi/cc)

H - 3 8.47 E-08 Ar-41 5.95 E-08 Cr 51 1.54 E-13 Mn-54 2.35 B-14 Co-58 1.12 E-14 Fe-59 4.94 E-14 Co-60 1.58 E-14 Zn-65 3.11 E-14 Kr 85 7.90 E-06 Kr 85m 1.93 E-08 Kr 87 4.49 E-08 Kr-88 1.09 B-07 Sr-89 1.00 E-15 Sr-90 1.00 E-15 Nb-95 1.62 B-14 Mo-99 3.44 E-13 RQ-103 8.01 E-15 I -131 2.76 E-14 Xe-131m 7.30 E-07 I -133 5.35 E-13 Xe-133 6.08 E-08 Xe-133m 1.77 E-07 Cs-134 8.52 B-15 I -135 1.22 E-09 Xe-135 1.03 E-08 Xe-135m 1.27 E-07 Cs-137 1.52 E-14 Xe-138 2.60 E-07 Ba/La-140 7.08 E-14 Ce-141 1.54 E-14 Ce-144 6.77 E-14 Gross Alpha 2.61 B-15 rt I" jl "P k) ~ )I II I'N b I 'h h 'f k 0 Semiannual Radioactive Bffluent Release Report January 1, 1987 to June 30, 1987 Appendix 2: Effluent and Haste Disposal Report Enclosure 1 : LOWER LIMITS OF DETECTION (LLD)

2. LLD's for Liquid Effluents NUCLIDE LLD(uCi/ml)

H 3 4.64 B-06 Na-24 3.28 E-08 Cr-51 1.59 E-07 Mn-54 2.14 B-08 Co-58 2.78 B-08 Fe-59 6.71 E-08 Co-60 3.85 E-08 Zn-65 1.07 E-07 Kr-85m 3.08 E-08 Sr-89 5.48 B-09 Sr 90 3.30 E-09 Zr 95 5.05 E-08 Nb-95 4.89 B-08 Mo-99 2.38 E-07 Tc-99m 2.73 E-08 Rh-105 1.16 E-07 Ru-105 8.44 E-08 I -131 3.07 E-08 I -133 3.35 E-08 Xe-133 8.74 B-08 Xe-135 2.57 E-08 Cs-134 2.68 E-08 Cs-137 3.80 B-08 Ba/La-140 1.17 E-07 Ce-141 3.87 E-08 Ce-144 2.00 E-07 H-187 8.91 E-08 Gross Alpha 5.85 E-08 ( W ir,,W W-.(> (8 )g( II P P>> flw, (( <<l (, "W, I" th), Jjihhhp II II I ~ h W hhll ),( ) ~ r: II W h') W F h ~ p~ ) Wh ff) If ) ('() "JW (h ()(.' ,,'gO h P ( r)h )l0 (W w W() }ay,," < '.W., t W',";. l hh ( ( ">',( i g-,"WI g,"jj H ()' ~ ~ t I I ) If h (Wh W( hl(( i 'will Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 2 : Effluent and Waste Disposal Report Enclosure 2 : Effluents Released Table lA GASEOUS EFFLUENTS SUMMATION OF ALL RELEASES Units Quarter Quarter Est. Total 1 2 Error A. Fission h Activation Gase 1.Total Release Ci 1.15E+02 6.33E+02 4.50E+Ol 2.Avg. Release Hate for Period uCi/sec 1.49E+Ol 8.05E+Ol 3.Percent of Tech. Spec. Limit 1.10E-Ol 3.00E-Ol B. Iodines 1.Total I-131 Ci O.OOE+00 O.OOE+00 2.00E+Ol 2.Avg. Release Rate for Period uCi/sec O.OOE+00 O.OOE+00 3.Percent of Tech. Spec. Limit O.OOE+00 O.OOE+00 C. Particulates 1.Particulates with Tl/2> 8 days Ci O.OOE+00 O.OOE+00 2.00E+Ol 2.Avg. Release Rate for Period uCi/sec O.OOE+00 O.OOE+00 3.Percent of Tech. Spec. Limit O.OOE+00 O.OOE+00 4.Gross Alpha Radioactivity Ci 2.14E-06 6.30E-07 D. Tritium 1.Total Release Ci O.OOE+00 O.OOE+00 3.00E+Ol 2.Avg. Release Rate for Period uCi/sec O.OOE+00 O.OOE+00 3.Percent of Tech. Spec. Limit O.OOE+00 O.OOE+00 2/3 y Iml>> ffm, >>lllf m,, 'lf '(f ft.) m mn >>;,I "'" I" .f I Ilf tl5 I ~ I I "f ',hf.F,f),)'I (<<'>> (", ~ a f '>>if il, "m 'f().,",, a, ffm ((h j ~ Ii ( ' ~ l ~ I f 'if" "hf)<<> ();>> ff'(> j'Tm)(),0 ~ ",'I ") f>>>, I f <) ',f.')().!) <<), () ",(1 h",((f ('" f (fS) m('l (.'0 f 'ffa r ) I I f gm ), m,If>>11 ~ ', >>, Ih(>>I f ~ %l 'f C))il m II 'f h pl h m I'), I ) I fh () f) ,i': f.', jjh ff f f "f ' "f . If' (f ff ~ I) fl() f ')II) i "') ~ ,f >>,ff, I'l' >><f j, ffll 't f fl f I '>>I'* <<Ij ~II, ~ i ) off.", ~ .f f ) ~ ',' ,fq f l'h (I(h ~ () () 'f I (" f) II) <)(,' )f)f) 4fm ff Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 2  : Effluent and Waste Disposal Report Enclosure 2  : Effluents Released Table 1B : GASEOUS BFFLUENTS ELEVATED RELEASES All releases at Shearon Harris are made as ground releases. 2/4 'W W Jl Il,l', W WW h'lh II ll W I ~ If IW h'I I ih N h I II IIW ' Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 2  : Bffluent and Waste Disposal Report Bnclosure 2  : Effluents Released Table 1C : GASEOUS EFFLUENTS GROUND LEVEL RELEASE Continuous Mode Batch Mode Nuclides Released Units Quarter 1 Quarter 2 Quarter 1 Quarter 2

1. Fission Gases H-3 Ci LLD LLD LLD NO BATCH Ar-41 Ci 1.308-03 2.008-03 1.338-03 RELEASES WERE Kr-85 Ci LLD I LD LLD MADB IN Kr-85m Ci 2.108+00 2.V08-02 LLD QUARTER 2

Kr-87 Ci LLD V.008-03 LLD Kr-88 Ci LLD 3.508-02 LLD Xe-131m Ci LLD LLD LLD Xe-133 Ci 8.508+01 6.288+02 7.778-03 Xe-133m 1.608-02 LLD Xe-135 Ci 2.808+01 4.468+00 9.728-06 Xe-135m Ci LLD LLD LLD Xe-138 Ci LLD LLD LLD Total for Period Ci 1.158+02 6.338+02 9.118-03

2. Iodines I-131 Ci LLD LLD LLD NO BATCH I-133 Ci LLD LLD LLD RELEASES WERE I-135 Ci LLD LLD LLD MADE IN QUARTER 2

Total for Period Ci LLD LLD LLD 2/5 'I <<II<< , ~ l I ~ hl Ih f'h Il<< ,Ilh(<< UM IU,/ ( U,.UKllf M Ij" M.~l" I'( <<II,I)gal~<<l f)Ill; [$ << I)J h<<>> t<< ~ I<<'I <<( 'l I "," ( <<h ,,>8":,~ >'(Ml ( l)II<<l Ih I )<<CP ' I jj I <<U f I li, II I<<il if l'l>> II() I l I" I "p<<(hajj r,, I IU jj<<v li., r f h <<f jj<<I, 'I Mjl'I,'i, CU l <> 'i I l i'<<g M S f () h, ()() I, fljj) <<<<]j I<<<<<< ~ >> I U la U.M <<M~, I U 'I ~MM ') ",'(II '()'<<, ( I' I U<<MM m~ f I UM ( U ~ I~ I <<I Ml fI i') Ij ~ ! << IM II" I( g I) i, M*") I l; IJ Ug' fl l<<Il M<<III I' I Ii h U ~ Ilh.. << ~ I 1 I i Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 2 : Effluent and Waste Disposal Report Enclosure 2 : Effluents Released Table 1C (Continued) GASEOUS EFFLUENTS GROUND LEVEL RELEASE Continuous Mode Batch Mode Nuclides Released Units Quarter 1 Quarter 2 Quarter 1 Quarter 2

3. Particulates Mn-54 Ci LLD LLD LLD NO BATCH Fe-59 Ci LLD LLD LLD RELEASES WERE Co-58 Ci LLD LLD LLD MADE IN Co-60 Ci LLD LLD LLD QUARTER 2

Zn-65 Ci LLD LLD LLD Sr-89 Ci LLD LLD LLD Sr-90 Ci LLD LLD LLD Mo-99 Ci LLD LLD LLD s 134 LLD Cs-137 LLD LLD LLD B a/La-140 Ci LLD LLD LLD Ce-141 Ci LLD LLD LLD Ce-144 Ci LLD LLD LLD Gross Alpha Ci 2.14 E-06 6.30 E-07 LLD Total for Period Ci 2. 14 E-06 6. 30 E-07 < LLD 2/6 <> I ,~r~ ~ I'> I l4I V y TqV ft 'I fVT'tT>> ' 'I.'t lr'7 .,l "1 7.f ff~~~f .".I 'I ... > I"'>'" ) I) '"P Itf' 'j ff'tl f t " 1' I Nfffff) Uf) TiI l't t I,/' fT,!f1 / T tT - ~ 7" " 7 i II t l e,> fl, ~ . ~ I I TT 'f..a I

5) "II' ) .

r ~ '1,f tt tt

t. ~

f8 II T T 7 j "') 7.T>>'

~ I ) ff1 Il qf A f7 '3 "IIgll lt at r ~ll 71,l I t tt', fll fj 1 fI I II f1 I I I Il ltd~ II "j, I< f 7tt It ll C.'t" i,f.f '.) tl "I tt,li, I/,ff 'p ~ ) tt (f,f Tt, I I'. II,',T.f tT ~t't 77 Tf $ g "I (I i '[ ~ 'I f1 I 'I 'I f~f $t IITj 7 j; t lt I 0,,,tlf~ I f II 4' tl s ""tt i Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 2  : Effluent and Waste Disposal Report Enclosure 2  : Effluents Released Table 2A LIQUID EFFLUENTS SUMMATION OF ALL RELEASES Units Quarter Quarter Est. Total 1 2 Error A. Fission h Activation Products 1.Total Release (not including Ci 3.21 E-02 6.38 E-02 3.50 E+Ol tritium, gases, alpha) 2.Average Diluted Concentration uCi/ml 9.45 E-08 6.69 E-08 During Period 3.Percent of Applicable Limit 1.51 E+00 4.58 E-Ol B. Tritium 1.Total Release Ci 2.45 E+00 5.79 E+Ol 3.50 E+Ol 2.Average Diluted Concentration uCi/ml 7.19 E-06 6.07 E-05 During Period 3 cent of Applicable Limit 2.40 E-Ol 2.02 E+00 C. Dissolved and Entrained Gases 1.Total Release (not including Ci 1.28 E-03 1.35 E-02 3.50 E+Ol tritium, gases, alpha) 2.Average Diluted Concentration uCi/ml 3.76 E-09 1.42 E-08 During Period 3.Percent of Applicable Limit 1.88 E-03 7.10 E-03 D. Gross Alpha Radioactivity 1.Total Release Ci LLD 2.73 E-04 3.50 E+Ol E. Volume of water released liters 1.54 E+07 2.11 E+07 1.00 E+Ol (prior to dilution) F. Volume of dilution water liters 3.25 E+08 9.33 E+08 1.00 E+Ol sed during period 2/7 I>> v<<I) Ii "<< It I. << J rf<< 1 I hdl 4$ <<)3) [ '"'VI) -' Ili If 1 '<<ft,>> i, -.1<< i l ')f l" It'I t <<g Ii> <><<<<" )' << << ~ p ) <<<<<< 1 ,1 ~ /hrh . <<<< ~ If h )' 1 <<t I '<<) II $~ f II ) ). ) 1', lt <<'<<l- << f << ) f '<>r 1 I ~ 1 $ <<I g gh<<VV )))<<h I )h ~ ft )h) << 'h ,I llh)$ << <<Il t I>> <<<< ~ '<<1<<1<< )I II <<,'l t " <<'f V I f (<<<<II L' '<<,'h 1 dt'h) p j ) i') 'I ~ <<"'I'h $ ~ )'.) h' h)11))(t 'Jl )>>di f j<<i<<<< ,,E ) 1<< 0SI I li', .' L'.l lf)"-I <<t f ' ~ t,) <<."I), <<V<<<<, I <<1(t g h'\ ~ Vvi << I <<1 Ill << <<ihr)<<),'I t <

> QP N<< ) f <<III)'<< j<<<< h )h,<<>> t ti P'.hht g) / I )f fr<< ) h ) r)<< )<<<<<< I i I Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 2  : Effluent and Waste Disposal Report Enclosure 2  : Effluents Released Table 2B LIQUID EFFLUENTS Continuous Mode Batch Mode Nuclides Released Units Quarter 1 Quarter 2 Quarter 1 Quarter 2 H-3 Ci 2.45 8+00 5.79 8+01 NO CONTINUOUS Na-24 Ci RELEASES WERE MADE IN 1.15 8-03 1.85 8-03 THIS PERIOD Cr-51 Ci 3. 78 8-03 2.21 8-03 Mn-54 Ci 1.03 8-03 1.09 8-02 Co-58 Ci 2.05 8-02 4.46 8-02 Fe-59 Ci 4.33 8-04 1.41 8-04 Co-60 Ci 2.86 8-04 1.02 E-03 Zn-65 Ci LLD LLD Sr-89 Ci LLD LLD Sr-SO Ci LLD LLD Zr/Nb-95 Ci 7.62 8-04 9.14 8-04 Mo-S9 Ci LLD LLD Tc-99m Ci 7.21 8-04 3.75 E-04 Rh-105 Ci LLD 4.49 8-05 Ru-105 Ci LLD 8.10 8-05 I-131 Ci 9.94 8-04 8.65 8-04 I-133 Ci l. 28 8-03 7.68 8-04 Cs-134 LLD LLD Cs-137 Ci LLD 2.00 8-05 Ba/La-140 Ci 1.06 8-03 LLD Ce-141 Ci LLD LLD W-187 Ci 1. 50 8-04 LLD Gross Alpha Ci LLD 2.73 8-04 Total for Period Ci 2.48 8+00 5.80 8+01 2/8 r)mr>J f fttrftlf,f K'I 'm m ~ ) mr m m Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 2 : Effluent and Waste Disposal Report Enclosure 2 : Effluents Released Table 2B (Continued) LIQUID EFFLUENTS Continuous Mode Batch Mode Nuclides Released Units Quarter 1 Quarter 2 Quarter 1 Quarter 2 Ar-41 Ci LLD LLD NO CONTINUOUS Kr-85m Ci LLD 1.75 E-05 RELEASES WERE MADE IN Xe-133 Ci 1.55 E-04 < LLD THIS PERIOD Xe-135 Ci 1.12 E-03 1.35 E-02 Total for Period Ci 1.28 E-03 1.35 E-02 2/9 W t W W VI IV ~ Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 2: Effluent and Waste Disposal Report Enclosure 3 : Solid Waste Disposal No radioactive waste or irradiated fuel was shipped during this report period. 2/10 A l W ,q g I'+ II e I I rl 4 II Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 3  : Changes to Process Control Program (PCP) Technical Specification 6.13 No changes were made to the PCP during this report period. lc tl k I' h h ',e)i) 1',<, g < > ~ J V Semiannual Radioactive Effluent Release Report January 1, 1987 to June.30, 1987 .Appendix 4: ~ Changes to the Off-Site Dose Calculation Manual (ODCM) Technical Specification 6.14 The following changes were made to the ODCM during the report period and during earlier plant start-up. Exhibit 1 provides a chronology of ODCM changes. Exhibit 2 provides a cross index of effective page changes. This exhibit identifies change locations in Revision 0.0 vs. Revision 1.0. Exhibit 3 provides .a listing of the ODCM changes. I Exhibit 4 presents the actual'changed 'pages of the ODCM. Change bars identify affected ar'eas,and.a,change,.number is ..given-for crossreference'.to those used in this appendix. 4/1 MEM/ATTACH4/OS2 EXHIBIT 1 'HRONOLOGY OF ODCM CHANGES The ODCM, Version 0.0, was approved by the Plant Nuclear Safety Committee (PNSC) on August 17, 1984. This version was submitted to the NRC on August 31, 1984. On April 4, 1985, the NRC requested four points of information. Three of these points required changes to the ODCM (see Change Items 22, 32, and 45). CPSL responded to the NRC information request on July 1, 1985. Version 0.0 of the manual was approved by the NRC together with the July 1, 1985 response, on May, 30, 1986.. Tentative changes to the ODCM were submitted for PNSC revie~ on August 8, 1985 and October 16, 1985. These, included Change Items: 1-53. Approval for these changes was requested of the PNSC on September 17, 1986 after .receiving formal approval .of, Version 0.0 from .'the .NRC.,'-'The PNSC =approved=these changes 'September 26', 1986. The new version of .the 'manual'.was designated'Revision 1.0, Draft 81. The Technical Specifications were issued together with the Low Power Testing License on October 24, 1986. Approval for Change Items 54 and 55 to the ODCM was granted by the PNSC on November 21, 1986. Approval for Change Items 56 through 60 to the ODCM was granted by the PNSC on June 3, 1987. 4/2 MEM/ATTACH4/OS2 EXHIBIT 2 CROSS INDEX OF EFFECTIVE PAGE CHANGES REVISION 0.0 VS. REVISION 1.0 CHANGE REV IS ION 0.0 REVISION 1.0 NUMBER PAGE NUMBER PAGE NUMBER 1-1 1-1 l-l 1 2 1-1 3 2-1 2-1 4 2-1 2-1 5 2-1 2-1 to 2-3 6 2-3 2-4 7 2-3 2-4 8 2-3 2-5 9 2-4 2-5 10 2-4 2-5 11 2-4 2-5 12 2-4 2-5 13 2-4 2-5 14 2-4 2-6 15 2-4 2-6 16 2-4 2-9 17 2-5 2-10 18 . '2-6 2-11 19 2-7 2-12 20 2-7 2-12 21 2"7 2-12 22 2-7 2-13 to 2-14 23 2-7 2-14 to 2"15 24 2-8 2-15 25 2-10 2-18 26 2-18 2>>18 27 2-11 2-19 28 2-12 2-19 29 2-12 2-20 30 2-13 2-21 31 2-18 2-26 32 NEW 2-27 33 NEW 2-28 34 3-4 3-4 35 3-4 3-4 36 3-5 t 3-5 o 3-6 37 3-8 3-8 38 3-8 3-8 39 3-8 3-9 to 3-13 40 3-8 3-13 41 3-9 3-15 42 3" 12 3-18 43 3-15 3-21 44 3-22 3-29 45 3-25 3-32 46 3"27 3-35 47 3-47 3-55 48 3-48 3-56 49 3-49 3-57 50 4-16 to 4-18 4-16 to 4-18 51 4-19 4-19 52 NEW 7-1 53 0-1 D-l 54 2-4 2-'4 55 2-5 2-5 56 2-3 2-4 57 2-4 2-5 58 2-4 2-6 to 2-9 59 2-13 2-21 60 2-18 2-26 4/3 MEM/ATTACH4/OS2 , EXHIBIT 3 ODCM CHANGES Page 1-1, Section 1.0, reference to Technical Specification "3.11.3" was deleted as the ODCM does not address Solid Radioactive Wastes. 2 ~ Page 1-1,Section 1.0, explicit mention is made for inclusion of "non-routine" releases in cumulative dose accountability to comply with 10CFR50 limits. 3 ~ Page 2-1, Section 2.0 now provides explicit discussion on the nature of potential non-routine liquid releases from the plant. 4.. Page 2-1, Section 2.1.1, the following two sentences have been deleted: "The blowdown flow rate, "B" is determined by the cooling tower basin water level. This water level-,.is adjusted depending, on the conductivity of the basin .water". "The sentences were 'deleted due to their -specificity, i.e , other operational parameters also legitimately influence blowdown from the Cooling Tower. 5 ~ Page 2-2, Section 2.1.1.la, subsection "a" is new. It is included to comply with footnote 2 of Table 4.11-1, Technical Specification 4.11.1 1 1 ~ ~ ~

6. Page 2-3, Section 2.1.1.lb, Equation 2.1-2, the term "n" has replaced the factor "10", where n is greater than or equal to 2. Using conservatism factors in set point calculations is at the option of the plant (NUREG-0133). Replacing the "hard and fast" factor of 10 with a selectable value provides greater flexibility in radwaste release operations.

7 ~ Page 2-4, Section 2.1.1.lc, at the definition of "B", the phrase "nominally, or estimated available flow rate" has,been added for clarification.

8. Page 2-4,'ection 2.1.l.lc, at the definition of, DFB; the definition has been made consistent with change (6) above.

9. Page 2-5, Section 2.1.1.ld, above Equation 2.1-4, the phrase "Determine monitor count rate above background:"

has been added for clarification.

10. Page 2-5, Section 2.1.l.d, at the definition of "CR",

the dimensions "cps" have been changed to "cpm" to be consistent with Radiation Monitor System (RMS) usage. 4/4 MEM/ATTACH4/OS2 Appendix 4: CHANGES (continued) Page 2-5, Section,2.1.l.ld, at, the definition of Em, the dimensions "cps/pCi/ml" have been changed to "cpm/pCi/ml" to be consistent with RMS usage.

12. Page 2-5, Section 2.1.1.1d, above Equation 2.1-5, the phrase "Determine monitor set point:" has been added for clarity.

13. Page 2-5, Section 2.1.1.1d, Equation 2.1-5 is new and permits calculation of the liquid radiation monitor set point in units of pCi/ml.

14. Page 2-6, Section 2.1.1.1d, at the definition of "CR",

the dimensions of "cps" are changed to "cpm" to be consistent with RMS usage.

15. -

Page 2-6, Section 2.1.1.1d, at the definition of "Bkg", the dimensions of "cps" are changed to "cpm" to be , consistent with .RMS,.usage.

16. Page 2-9, Section 2.1.1.1e,,Equation 2.1-6 replaces the term MRR (i.e... Maximum Release Rate) with the term RR (i.e. , the anticipated Release Rate) where the RR should not exceed the MRR see the definition of RR, which is also new. Use of RR permits greater flexibility in radwaste release operations.

In addition, the RR term is also included in the denominator. Inclusion of the term is appropriate pursuant to NUREG-0133.

17. Page 2-10, Section 2.1.l.le, at the definition of "B",

the phrase "nominally, or, estimated available flow rate" has been added for clarity.

18. 2-11, Section 2.1.1.2a, at Equation 2.1-8, the term Page "Vk" (the 'k's a subscript), is now included in the denominator. Inclusion of the term is appropriate pursuant.to NUREG-0133.

19. Page 2-12,'Section '2.1.1.2b, a ."Note" quotes the 10CFR20 criteria for determination of radioactivity in a sample mixture.

20. Page 2-12, Section 2.1.2, the word "monthly" has been replaced by the word "weekly" pursuant to the FINAL DRAFT of Technical Specification Table 4.11-1.

21. Page 2-12, Section 2.1.2, the phrase "(see note in Section 2.1.1.2b)" is added for clarification.

4/5 MEM/ATTACH4/OS2 Appendix 4: CHANGES (continued)

22. Pages 2-13 and 2-14, Section 2;1.2.1 entitled "Set points for the Normal,.service Water (NSW) Monitors" is new and describes the set point methodologies for these monitors. This methodology was requested by the NRC (reference letter, S.R. Zimmerman to H. R. Denton, July 1, 1985, NLS-85-226).

23. Pages 2-14 and 2-15, Section 2.1.3 entitled "Non-routine Liquid Releases" provides detailed discussion of non-routine liquid effluent release situations at the plant.

24. Page 2-15, Section 2.2.1, the phrase " ~ . . and all defined periods of continuous release . . ." has been added for clarity.

25. Page 2-18, a paragraph had been added explaining the conservatism in, including the Lillington Municipal Water Facility as a drinking water pathway for the plant. The paragraph is reproduced below and was in response to a

-,technical;-specification:that',did"not become-a part of the final specifications. Because of this, the paragraph was eventually deleted as unnecessary. Inclusion of the drinking water pathway for SHNPP is conservative since the Lillington Municipal Water Facility is located at a point greater than three miles from the plant (see, footnote in Technical Specification 3.11.1.2, Action a).

26. Page 2-18, Section 2.2.1, the words " ...receptor...

...locale..." have been added for clarity.

27. Page 2-19, Section 2.2.1, the sentence beginning with:

"This report ..." has been corrected grammatically.

28. Page 2-19, Section 2.2.2, Equation 2.2-8 provides the dose projection formula for liquid effluents.

29. Page 2-20, Section 2.2.2, Equations 2.2-9 and 2.2-10 give 'the dose projection limits for liquid effluents.

30. Page 2-21, Table 2.1-1,Eductor factors for liquid effluent release tank have been included in support of the mixing methodology; see Change Item (4) above. The table has also been reformatted for better presentation. Finally,= a 100 gpm value has been added for REM-3540 recirculation flow rate.

31. Page 2-26, Figure 2.1-2, the "Settling Basin" is now shown in order to depict the effluent pathway more accurately.

4/6 MEM/ATTACH4/OS2 Appendix 4: CHANGES (continued) ,32 .Page 2-27, Figure 2.1-3, the Normal Service Water Flow diagram is new and is, in response to a NRC request to have such a diagram included in the ODCM (reference letter S; R. Zimmerman to H. R. Denton, July 1, 1985, NLS-85-226) ~

33. Page 2-28, Figure 2.1-4, the "Other Liquid Effluent Pathways" diagram shown in this figure is new and shows the possible non-routine liquid effluent lines from the plant.

34. Page 3-4, Section 3.1.1.4, the term "f" is now summed into the denominator. The term is included to account more explicitly for significant addit'ional vent stack flow due to batch releases. Inclusion of the term is conservative inasmuch as it lowers the set point value.

35. Page 3-4, Section 3.1.1.4, a "Note" is included that references the, FSAR .chapter, where the design basis, vent stack'flow;.rates",can'be found.

36. Page 3-5, Section 3.1 1.6, a "Note" is included to

~ explain how gaseous effluent monitor set points can be converted to dimensions of pCi/sec.

37. Page 3-8, Section 3.1.2.2, same as Item 34 earlier.

38. Page 3-8, Section 3.1.2.2, at the definition of "F" the phrase " . . . or the actual flow rate" is added for clarity and operational flexibility.

39. Pages 3-9 through 3-13, Section 3.1.3, provides an additional alternative set point determination method for batch gaseous releases from the plant.

40. Page 3-13, Section 3.1.4, provides the following discussion for effluent monitoring during hogging operations.

If the, reactor has been, shut. down for less" than 30 days, the condenser vacuum discharge during initial hogging operations at plant start-up and prior to turbine operation will be routed directly to Turbine Building Vent Stack 3a. In this event, the set point methodologies of Sections 3.1.1 and 3.1.2 for the noble gas monitor located on Vent Stack 3a (see Appendix D) are applicable. 4/7 MEM/ATTACH4/OS2 'r N Appendix 4: CHANGES (continued) If the reactor has been shut down for greater than 30 days, the. condenser vacuum pump. discharge'uring .initial hogging operations at plant start-up and,prior to turbine operation may be routed as dual exhaust to (1) the Turbine Vent Stack 3a and (2) the atmosphere directly. In this instance, the blind flange on the latter exhaust route will be removed (see Figure 3.3). Set point determination in this case depends on knowledge of the flow rates through each of the exhaust pathways. Once these flows are established or estimated, the ratio of the flow through Vent Stack 3a to the flow in the direct exhaust path will be computed. . This ratio will be used to reduce the set point on Vent Stack 3a to account for noble gases being exhausted concurrently via dual pathways. [END] The discussion is provided persuant to close out of ,. Safety. Evaluation Report open Item No. 9. 41'. Page .3-15, Table 3.1-1, typographic correction. The values 9.44E 01 and 2.23E 02 were corrected to 9.44E-01 and 2.23E-02, respectively, at Si column under Containment Purge or Pressure Relief via Vent Stack 1.

42. Page 3-18, Section 3.2.1, the sentence: "Table 3.2-2 presents the distances from SHNPP to the nearest area for each of the 16 sectors as well as to the nearest residence, vegetable garden, cow, goat, and meat animal." has been deleted as unnecessary.

43. Page 3-21, Section 3.2.2, a new paragraph was created at "However . . ." for editorial clarity. This involved no text deletion or addition.

44. Page 3-29, Section 3.3.1'.2, Equations 3.3-7 through 3.3-9 give the dose projection formula and dose limits for noble gases in gaseous effluents.

45.;. Page 3-32, Section 3.3.2 ', 'at. definition of RiB, typographic correction.,changing the word "vegetable" to meat

46. -

Page 3-35, Section 3.3.2.2, Equations 3.3-13 and 3.3-14 provide the dose projection formula and dose limit for particulates and radioiodines in gaseous effluents. 4/8 MEM/ATTACH4/OS2 Appendix 4: CHANGES (continued)

47. Page 3-55, Figure 3.1,;the containment pre-entry purge influent line monitor to the plant vent is now labeled with its identification number. Also, the presence of the Wide Range Gas Monitors is now identified and labelled appropriately.

48. Page 3-56, Figure 3-2, has been improved and corrected. The location of Vent Stack 3a is now in the appropriate position on the Turbine Building.

49 'age 3-57, Figure 3.3 has, been improved and updated. The diagram now shows the presence of (1) the Wide Range Gas Monitor, (2) the removable blind flange on the hogging line and (3) proper placement of the gland steam condenser influent to Vent Stack 3A.

50. Page 4-16. through 4-18, Figures 4.1-2 through 4.1-4 have been improved.

51 ' 'Page'4-19,,Figure 4.1-5,'.has been" corrected with addition of "bottom sediment" and "shoreline sediment" sample designations.

52. Page 7-1, Section 7.0 entitled: "Licensee-Initiated Changes to the ODCM" has been added for explanatory purposes and regulatory reference.

53 'age effluent monitors on D-l, Appendix D, now lists the non-routine pathway outdoor tank area drain (1) the transfer monitor line and (2) the turbine building floor drains effluent line. Also, the Normal Service Water (NSW) monitors are listed as well as the Wide Range Gas Monitors (WRGMs) and the Containment Pre-entry Purge line monitor. Most of these monitors are included for information only.

54. Page 2-4, Section 2.l.l.lc. For clarity, include the following.'OTE'his method of determining the Maximum Release Rate (MRR) ensures. conformance with the,.test in Section F below.

55. Page 2-5, Section 2.1.1.1d, Equation 2.1-5 at the definition of SPc. Previous definition read: SPc = 2CR

+ Bkg. Revision of definition would read: SPc = 2 (CR + Bkg). 4/9 MEM/ATTACH4/OS2 Appendix 4: CHANGES (continued) Where: SPC = liquid monitor set point; cpm CR = monitor count rate above background given by the summation of the radionuclidic concentration in the tank multiplied by the monitor efficiency', cpm Bkg = monitor background; cpm The revision is because when the CR value is 0.0 cpm, i.e., there is no radioactivity, this would set the liquid monitor set point to background incurring the possibility of spurious alarms due to background fluctuation. The revision would correct for this by doubling the observed background value and allowing this to be used for the.set point. This .".,engineering .factor" is like others used in 'the ODCM to prevent similar problems.

56. Page 2-4, Section 2.1.1.1c. Change Equation 2.1-3 from:

MRR = B to MRR = B (T ) 2(DFB) ~DFB Also include the following definition: Tm = Fraction of the radioactivity from the site that may be released via the monitored pathway to ensure that the site boundary limit is not exceeded due to simultaneous releases from more than one pathway. The Tm sum for the site shall not exceed one (1). And delete the following definition: Engineering factor to prevent spurious alarms caused by deviations in the mixtures of radionuclides which affect the monitor response. The change permits a more flexible determination of the MPC allocation for any given waste stream during concurrent release conditions.

57. Page 2-5, Section 2.1.1.1d, Equation 2.1-5 at the definition of SP . Change the following SP = 2(CR +

Bkg) to: SP = CR + Bkg + 3.3 ~gk 2T 4/10 MEM/ATTACH4/OS2 11 ',r, ~l f .. Appendix 4: ..CHANGES (continued) Also, replace the following definition'. 2 = Engineering factor to prevent spurious alarms .caused by deviations in the mixture of radionuclides which affect the monitor response (see determination of Equation 2.1-3). With' new definition.'. 3 ~Bk Statistical variance on the 2T background (Bkg) counting rate quoted at the 99.95X confidence level at a time constant v (min) which is a function of Bkg. This te'rm is included to prevent inadvertent high alarm trips due to ."random.;fluctuation:,in the monitor background. This change is made to account for radiation monitor background fluctuations more directly and with a known statistical confidence level.

58. Page 2-6 to 2-9, Section 2.1.1.1d. Include the following text providing two alternative methods of calculating the set point for liquid effluent radiation monitors.

ALTERNATIVE SET POINT METHOD BASED ON I 131 MPCW This method conservatively assumes: I. All of the radioactivity is due to I-131, which has the lowest Maximum Permissible Concentration (MPC ), persuant to 10CFR20. II. Only'the minimum cooling, tower blowdown flow rate is available for dilution. III. The maximum effluent discharge flow rate is utilized. Determine SP , the set point above background in pCi/ml. B + MRR (2.1-5A) = SPm MPC I 131 ( MRR ) (Tm) 4/11 MEM/ATTACH4/OS2 Appendix 4: CHANGES (continued) where.'SP = set point above background (pCi/ml) MRR = Maximum effluent discharge flow rate (gpm) B = Minimum dilution flow rate (gpm) T = Fraction of the radioactivity from the site that may be released via the monitored pathway to ensure that the site boundary limit is not exceeded due to simultaneous releases from more than one pathway.. The sum of T for the site shall not exceed one (1). Determine SP , the set point above background in cpm. c'= ( m) (m) (2.1-5B) where: SP = set point above background (cpm) SP = set point above background (pCi/ml) E = Monitor efficiency (cpm/pCi/ml) Add the monitor background to either SP or SP to determine the monitor setting for the high alarm set point. ALTERNATIVE SET POINT METHOD BASED ON ANALYSIS OF EFFLUENT PRIOR TO DISCHARGE This method provides a set point using a more precise evaluation which includes the actual cooling to~er dilution flow rate, effluent discharge flow rate and an analysis of the principal gamma emitters in the liquid effluent to be released. Determine SP , the set point above background in pCi/ml. SPm = g B (2.1-5C) where: SPm set point above background (pCi/ml) )c Total radioactivity concentration of gamma-emitting radionuclides in liquid effluent prior g to dilution (uCi/ml). Effluent discharge flow rate (gpm) Cooling tower blowdown flow rate (gpm) DFB given previously in equation 2.1-2. 4/12 MEM/ATTACH4/OS2 I Appendix 4: CHANGES (continued) Tm =,Fraction of the radioactivity,.from the site that may be released via the monitored pathway to ensure that the site boundary limit is not ,exceeded due to simultaneous releases from more than one pathway. The sum of T for the site shall not exceed one (1). Determine SPc the monitor set point above background in cpm. SP = (SP ) (E ) (2.1-5D) where: SP = set point above background (cpm) SPm = set point above background (pCi/ml) E = monitor efficiency (cpm/pCi/ml)

Add;,the monitor"background;,to;either",.SP .or SP to,determine -the monitor, setting. for'.the high alarm set point.

If it is determined that f + B the release can be made. If it is determined that f + B DFB f) the release cannot be made. Reevaluate the discharge flow rate prior to dilution and/or the dilution flow rate. The first alternative method (Eq. 2.1-5A) bases the set point on (1) the I-131 Maximum Permissible Concentration which is the lowest MPC> found in 10CFR20; (2).theminimum assured dilution flow'rate and (3) the maximum available effluent discharge flow rate. 'his method is expected to be useful once SHNPP achieves steady state operating conditions. The second alternative liquid set point method (Eq. 2.1-5C) utilizes the characteristics of each batch liquid release in setting the high alarm set point value. This approach will generate variable high alarm set points depending upon (1) the specific radionuclidic mix in the liquid effluent; (2) the available dilution flow rate and,(3) anticipated discharge flow rate for the release'. The method provides a flexible approach to set point determination that will facilitate optimization of dilution and discharge flow rates. 4/13 MEM/ATTACH4/OS2 Appendix 4: CHANGES (continued)

59. Page 2-21, Table 2.1-1, pump capacities for the SWST and TLEHS tanks have been correct'ed. Also the eductor factors have been updated from the pre-startup estimates given earlier to more realistic calculated values.

60. Page 2-26, Figure 2.1-2 has been corrected to indicate the separate influent point from the settling basin to the cooling tower blowdown line.

MISCELLANEOUS CHANGES In conformance to the final draft of the Technical Specifications references to "site, boundary" were changed to "exclusion boundary". The Table of Contents has been altered to reflect the presence of Chapter 7 and new pagination. 4/14 MEM/ATTACH4/OS2 EXHIBIT 4 CHANGED PAGES FROM THE ODCM 4/15 MEM/ATTACH4/OS2

1.0 INTRODUCTION

The Off-Site Dose Calculation Manual (ODCH) provides the information and meth-odologies to be used by Shearon Harris Nuclear Power Plant, (SHNPP) to ensure compliance with Specifications 3.11.1, 3.11.2, and 3.11.4 of the SHNPP Tech-nical Specifications. These portions are those related to normal liquid and gaseous radiological effluents. They are intended to show compliance with 10CFR20, 10CFR50.36a, Appendix I of 10CFR50, and 40CFR190 in terms of appro-pr i ate monitor ing instrumentation, dose rate, and cumulative, dose l imi-tations. Off-site dose estimates from nonroutine releases, wil al so be 1

included in the cumulative. dose estimates for the plant to comply wi.th Appendix I of IOCFR50.

The ODCH is based on "Westinghouse Standard Technical Specifications" (NUREG 0452), "Preparation, of Radiological Effluent Technical Specifications for Nu- .

clear Power Plants" (NUREG 0133), and guidance from the United States Nuclear Regulatory Commission (NRC). Specific plant procedures for implementation of this manual are presented, in the SHNPP Plant Operating Manual and other con-1 trolled documents. These procedures will be utilized by the operating staff of SHNPP to ensure compliance with technical specifications..

The ODCM has been prepared as generically as possible in order to minimize the need for future revisions. However, some changes to the ODCH are expected in the future. Any such changes will be properly reviewed and approved as indi-cated in the Administration Control Section Specification 6.14.2 of the SHNPP Technical Specifications.

ODCH (SHNPP) Rev. 1.0

2.0 LIQUID EFFLUENT Liquid releases at SHNPP are divided into batch and continuous modes. Each mode is further separated into routine and nonroutine release paths. Routine batch releases are expected via process streams described in Section 2.1.1.

Nonroutine batch releases are effluent paths that only have the potential for.

containing radioactivity. The outdoor tank 'area drain line, the turbine building floor drains effluent line (yard oil separator line), and the efflu-ent from from the secondary waste treatment system (SWTS) are considered as nonroutine batch release points. In the SWTS, this is true only when no radioactivity is detectable due to primary to secondary leakage. These efflu-ent paths are monitored for radioactivity (see Appendix D and Figures 2.1-2 and 2.1-4) and should the setpoint be exceeded, releases are automatically terminated. Further discussion of these effluent lines is provided in Sec-tion 2.1.3.

Planned continuous liquid releases containing radioactivity do not presently occur at SHNPP and thus these are considered as nonroutine release pathways.

Section 2.1.2 describes continuous releases in greater detail.

2.1 COMPLIANCE WITH 10CFR PART 20 (LIQUIDS) 2.1.1 Batch Releases A batch release'is the discharge of liquid waste of a discrete volume. Batch releases from the SHNPP liquid vadwaste system may occur from treated laundry and hot shower tanks, secondary waste treatment tank, waste monitor tanks, and waste evaporator condensate tanks. The principal sources'f waste for these tanks are shown in Figure 2.1-1.

The liquid radwaste effluent streams are shown in Figure 2.1-2. A batch release represents the emptying of one tank only. No concurrent liquid batch releases (i.e., more than one tank at a time) are made from SHNPP. The liquid radwaste system discharges to the cooling tower blowdown line. Dilution flow depends primarily on the blowdown Flow "B." If liquid effluent is diverted to the waste neutralization basin, some additional dilution may also occur at ODCM (SHNPP) 2-1 Rev. 1.0

thi s point. For the purpose of cal cul ation, the assumed value of B i s 16.5 cfs (7.4E3 GPM) as presented in the SHNPP FSAR, Section 11.2.3. This value is presently interpreted as the average blowdown flow rate but may be variable. If B is less than 16.5 cfs, then the measured flow rate should be used The sampling and analysis frequency and the type of analyses required by the SHNPP Technical Specifications are given in Table 4. 11-'1 of the specifica-tions. All applicable radiation monitoring instrument numbers are listed in Appendix D.

2.1.1.1 Prerelease The. radioactive content of each batch release will be determined prior to release in accordance with Table 4.11-1 of the SHNPP Technical Specifica-tions. Compliance with 10CFR20 will be shown in the following manner:

a. Mixing Method for Isolated Liquid Effluent Tanks Prior to Sam-pling for Radioactivity Analyses Equation 2.1-0 below provides an acceptable method for ensuring a well-mixed tank so that a representative sample can be taken for radioactivity or'ther appropriate analyses. The method addresses the requirement found in Foot-note 2, Table 4.11-1, of Technical Specification 4. 11. 1. 1. 1.

. (V) (E) (n) (2.1-0)

(P) (60) where:-

Estimated .mixing time, hr Tank volume, gal Eductor factor Pump recirculation flow rate, gpm ODCM (SHNPP) 2-2 Rev. 1.0

Number of tank volumes for turnover; this will be typically two or more

- 60 60 min/hr Table 2.1-1 lists the volumes, eductor factors, and pump recirculation flow rates for individual liquid effluent release tanks.

b. Minimum acceptable dilution factor:

where:

DFo l,l Z

C.

MPC, (2. 1-1)

DFo Minimum acceptable dilution factor determined from a gamma isotopic analysis of liquid effluent to be released Ci Concentration of radionu'elide "i" in the batch to be released, pCi/ml MPC Maximum permissible concentration of radio-nuclide "i" from Appendix B, Table II, Col-umn 2, of 10CFR20, pCi/ml DFB n (DFo) (2.1-2) where:

DFB Conservative dilution factor used by SHNPP to calculate maximum release rate prior to re-lease in order to ensure compliance with 10CFR20 ODCM (SHNPP) 2-3 Rev. 1.0

A factor of > 2; 10CFR20 limits as specified in Appendix 8, Table II, Column 2. This factor represents one layer of conservatism for all releases at SHNPP DFo Minimum acceptable dilution factor per Equa-tion 2.1-1

c. Maximum release rate:

B MRR ~Tm~ (2.1-3)

~OV B where:

MRR Maximum release rate of the batch to be re-leased, gpm Cooling tower blowdown flow rate, gpm 7.4 E3 gpm nominally or estimated available flow rate Tm Fraction of the radioactivity from 'the site

'that may- be released via monitored pathway to ensure that the site boundary limit's not exceeded due to simultaneous releases from more than one pathway. The T sum for the site shall. not exceed one (1)

DFB Minimum acceptable dilution factor (DFo) made conservative by a factor of "n" per Equation !OS'.

1-2 Note: This method of determining the Maximum Release Rate (MRR)

Q+~

ensures conformance with the test in Section F below.

ODCM (SHNPP) 2-4 Rev. 1.0

d. Monitor Alarm/Trip Setpoint:

Monitor alarm/trip setpoints are determined to ensure that the concentration of radionuclides in the liquid effluent released from the site to unrestricted areas does not exceed the limits specified in 10CFR20, Appendix B, Table II, Column 2, for radio-nuclides other than dissolved or entrained noble gases. An MPC of 2 E-4 pCi/ml been established for noble gases dissolved or entrained in liquid effluents, based on the assumption that xenon-135 is the controlling radionuclide.

Determine monitor count rate above background:

(E C)' E (2.1-4)

CR 1 1 where:

CR Calculated monitor count rate above back-Oo ground, cpm Ci Concentration of radionuclide "i" in the. batch to be released, yCi/ml Em The monitor ef fi ci ency for the mixture of radionuclides in the liquid effluent prior to dilution, cpm/uCi/ml Determine monitor setpoint:

SP c

SPm * (2. 1-5)

E m

where:

SPm Monitor alarm/trip setpoint, qCi/ml ODCM (SHNPP) 2-5 Rev. 1.0

Bkg SPc CR + Bkg + 3.3 2T Bkg 3%3 Statistical variance on the background (Bkg) 2T counting rate quoted at the 99.95K confidence level at a time constant < (min) which is a function of Bkg. This term is included to prevent inadvertent high alarm trips due to random fluctuation in the monitor background.

Calculated monitor count rate per Equa-CR tion 2.1-4, cpm I~

Bkg Background count rate due to internal contami-nation and the radiation levels in the area in which the monitor is installed when the de-tector sample chamber is filled with an uncon-taminated fluid, cpm I is I

CAUTION: This setpoint must be evaluated as conforming to the test of "Section f" below.

ALTERNATIVE SETPOINT METHOD BASED ON I-131 MPCw This method conservatively assumes:

(1) All of the radioactivity is due to I-131, which has the lowest Maximum Permissible Concentration (MPCw), persuant to 10CFR20.

(2) Only the minimum cooling tower blowdown flow rate is avail-able for dilution.

(3) The maximum effluent discharge flow rate is utilized.

Determine SPm, the setpoint above background in pCi/ml.

ODCM (SHNPP) 2-6 Rev. 1.0

B+ MRR S'm (2.1-5A) 1-131 MRR m where:

SPm Setpoint above background (qCi/ml)

MRR Maximum effluent discharge flow rate (gpm)

Minimum dilution flow rate (gpm)

Fraction of the radioactivity from the site that may be released .via the monitored pathway to ensure that the site boundary limit is not exceeded due to simultaneous releases from more than one pathway. ,The sum of T for the site shall not exceed one (1).

Determine SPc, the setpoint above background in cpm.

SP (SP.) (E.) (2.1-5B) where:

SPc Setpoint above background (cpm)

SPm Setpoint above background (pCi/ml)

Monitor eff iciency (cpm/pCi/ml)

Add the monitor background to either SPm or SPc to determine the monitor setting for the high alarm setpoint.

ALTERNATIVE SETPOINT METHOD BASED ON ANALYSIS OF EFFLUENT PRIOR TO DISCHARGE ODCM (SHNPP) 2-7 Rev. 1.0

This method provides a setpoint using a more precise evaluation which includes the actual cooling tower dilution flow rate, effluent dis-charge flow rate, and an analysis of the principal gamma emitters in the liquid effluent to the released.

Determine SPm, the setpoint above background in yCi/ml.

z C (f+B)

SPm ) (T) (2.1-5C) g (DF) (f) m where:

SPm Setpoint above background (pCi/ml) cg Total radioactivity concentration of gamma-emitting radionuclides in liquid effluent prior to dilution (pCi/ml).

Effluent discharge flow rate (gpm)

Cooling tower blowdown flow rate (gpm)

DFB Given previously in Equation 2. 1-2.

Tm Fraction of the radioactivity from the site that may be released via the monitored pathway

~

to ensure that the site boundary limit is not exceeded due to simultaneous releases from more than one pathway. The sum of Tm for the site shall not exceed one (1).

Determine SPc, the monitor setpoint above background in cpm.

SP (SP ) (E ) (2.1-5D) where:

ODCM (SHNPP) 2-8 Rev. 1.0

SPc Setpoint above background (cpm)

S'm Setpoint above background (uCi/ml)

'm Monitor efficiency (cpm/pCi/ml)

Add. the monitor background ,to either SPm or SPc to determine the monitor setting for the high alarm setpoint.

If it is determined that f+B ) 1 (DF )

B (f) the release can be. made.

If it is determined that f+B (1 (DFB) (f) the release cannot be made. Reevaluate the discharge flow rate prior to dilution and/or the dilution flow rate.

Calculated concentration at unrestricted area:

(C.) (RR)

Conci RR+B (2.1-6) where:

Conc. Calculated concentration of radionucl-ide "i" at the unrestricted area, yCi/ml Ci Concentration of radionuclide "i" in the batch to be released, uCi/ml ODCM (SHNPP) 2-9 Rev. 1.0

RR Anticipated release rate of the batch that should not exceed the MRR as per Equation 2.1-3e gpm Cooling tower blowdown flow rate, gpm 7.4 E3 gpm nominally, or estimated available C~

flow rate

f. 10CFR20 Prerelease Compliance Check:

Before initiating the batch release, perform one final check for compliance with 10CFR20. If the sum of the ratio of .liquid'con-centration to MPC for all radionuclides- at the unrestricted area is less than or equal to 1, then 10CFR Part 20 limits have been met. The following equation must be true:

z Conc./MPC. ( 1 i 1 1

where:

Conc> Calculated concentration of radionuclide "i" at the unrestricted area per Equation 2. 1-6,

. uCi/ml Maximum permi ssibl e concentration of radi o-nuclide "i" from Appendix B, Table II, Column 2, of 10CFR20, yCi/ml 2.1.1.2 Postrel ease The actual concentration of each radionucl'ide following a batch release from a tank will be calculated to show final compliance with 10CFR20 as follows:

a. Actual concentration at unrestricted area:

ODCM (SHNPP) 2-10 Rev. 1.0

(C.) (V)

Concik (2.1-8)

V + V where:

Conc; >

The actual concentration of radionuclide "i" at the unrestricted area during release "k,"

pCi/ml Ci Concentration of radionuclide "i" in the batch released, gCi/ml Actual volume of 1 i qui d ef fluent re 1 eased during release "k," gal (see Table 2.1-1 for waste tank volumes and pump capacities).

Vd Actual volume of dilution water during release "k," gal (B) (tk) where:

Cooling tower blowdown flow rate, gpm Dur'ation of release "k," min

b. 10CFR20 Postrel ease Compliance Check:

To show final compliance with 10CFR20, the following relationship

~

must hold:

~ ' ~

z (Concik /HPC.

1 where:

ODCH (SHNPP) 2-11 Rev. 1.0

Concik The actual concentration of radionuclide "i" during release "k" (from Equation 2.1-8),

gCi /ml MPCi Maximum permissible concentration of radio-nuclide "i" from Appendix B, Table IZ, Column 2, of 10CFR20, uCi/ml Note: Pursuant to 10CFR20 Appendix 8, Note 5, " . . . a radionuclide may be considered as not present in a mixture if (a) the ratio of the concentration of that radionuclide in the mixture (CA) to the concentration limit for that radionuclide specified in Table II of Appendix "B" (MPCA) does not exceed 1'/10 (i.e.,

CA/MPCA < 1/10) and (b) the sum of such ratios for all the radionuclides considered as not present in the mixture does not exceed 1/4, i.e., CA/MPCA + CB/MPCB . . . + < 1/4."

2.1.2 Continuous Releases A continuous release is the discharge of liquid wastes of a nondiscrete vol-ume; e.g., from a volume or system that has an input flow during the contin-uous release. Planned continuous releases do not presently occur at SHNPP, although the potential does exist in the Normal Service Water (NSW) System and Emergency Service Water (ESW) System. The returns from the NSW System to the Circulating -Water System are monitored by installed radiation monitors which are covered by Technical Specification 3.3.3.10. In addition, a weekly com-posite sample is collected and analyzed in accordance with Technical Specifi-cation Table 4.11-1. If radioactivity is detected in either system,.it will be eventually diluted by flow from the Circulating Water System. Thus, dilu-ted effluent concentrations can be either computed with knowledge -of the circulating water flow and/or monitored by periodic sampling of the Cooling Tower Basin. In the event radioactivity is detected in the Emergency Service Water System, then ESW flow, the Cooling Tower Basin, and the return flow to the auxiliary reservoir wil,l be periodically sampled. To show compliance with 10CFR20, the sum of the concentration of radionuclide "i" in the unrestricted area due to both continuous and batch releases divided by that isotope's MPC IC~P must again be less than 1 (see note in Section 2.1. 1.2b).

ODCM (SHNPP) 2-12 Rev. 1.0

2.1.2. 1 Setpoints for the Normal Service Water (NSW) Monitors C

Figure 2.1-3 is a diagram of the NSW system. A radiation monitor is located on each of the NSW returns to the circulating water system and they are indi-cated in the diagram. Either of two methods may be used to determine the setpoints for the NSW radiation monitors.

Method 1: Use Equation 2.1-10 below:

CPM bkg MOC = 2 (2.1-10) 2T Sensitivity where:

MDC Minimum detectable concentration for a given isotope or isotopic mix (pCi/ml) cpmbkg Ambient cpm + (mR/hrbk

• cpm/mR/hr) bkg Time constant of signal processor (min). This is a function of cpmbkg sensitivity = For selected isotope or isotopic mix (cpm/

.pCi/ml)

Method 2: Use Equation 2.1-11 below:

SP c

SPm (2.1-11)

E where:

SPm Setpoint, pCi/ml 0OCM (SWPP) 2-13 Rev. 1.0

SPc (2) (bkg); cpm Engineering factor to account for spurious

. alarms Em The monitor efficiency for the mixture of radionuclides in the liquid effluent (cpm/

gCi/ml) bkg Background count rate due to internal radia-tion levels in the area in which the monitor is installed when the detector views an uncon-taminated fluid (cpm)

Method 2 is acceptable from an effluent release standpoint because HSW is not discharged directly to the environment and it undergoes significant dilution in the cooling tower basin.

2.1.3 Nonroutine Liquid Releases 2.1.3.1 Outdoor Tank Area Drain Effluent Line The outdoor tank area drain effluent line routes rainwater collected in the outdoor tank area to the storm drain system and from there to the cooling tower blowdown line for release to the environment. The line is monitored for radioactivity and is capable of automatic termination of effluent release.

Because no radioactivity is normally .expected in this line, the monitor set-point can be -determined with either Equation 2.1-10 or 2.1-11. If 'the set-point is exceeded, the release is automatically terminated. Effluent can then be diverted to the floor drain system for processing and eventual release via the waste monitor tanks (see Figures 2.1-1 and 2.1-2).

2.1.3.2 Turbine Building Floor Drains Effluent Line Water collected in the turbine building floor drains is normally routed to the w

yard oil separator for release to the environment via the waste neutralization ODCM (SHNPP) 2-14 Rev. 1.0

system and then to the cooling tower blowdown line. Because no radioactivity is normally expected in this path, the setpoint for the radioactivity can be determined with either Equation 2.1-10 or 2.1-11. Should the setpoint be exceeded, the release is automatically terminated. Effluent can then be diverted to the secondary waste treatment system for processing and eventual release via the secondary waste treatment tank (see Figures 2.1.1 and 2. 1-2).

2.1.3.'3 Secondary Waste Treatment System (SWTS)

When no radioactivity is detectable due to primary to secondary leakage, effluent from the SWTS may be released directly to the environment. In this event, the setpoint for the radioactivity monitor can be determined with either Equation 2.1. 10 or 2.1.11. Should the setpoint be exceeded, the re-lease is automatically terminated.

2.2 COMPLIANCE WITH 10CFR50 2.2.1 Cumulation of Doses The dose contribution from the release of liquid effluents will be calculated at least once every 31 days (monthly), and a cumulative summation of these total body and any organ doses will be maintained for each calendar quarter.

The dose contribution for batch releases and all defined periods of continuous release will be calculated using the following equation:

-z,.t 0 lv k ik k ) (2.2-1)

.k i )

where:

D The cumulative dose commitment to the total .

body or any organ ~, from the liquid effluents releases, mrem:

ODCM (SHNPP) 2-15 Rev. 1.0

730 Adult water consumption rate (from Table E-5 of Regulatory Guide 1.109) Rev. 1, liters/yr.

Dw Dilution factor from the near-field area within one-quarter, mile of the release point to the potab1 e water intake for the adul t water consumption 13.95 for uptake at the municipal water faci 1-ity at Lillington BF Bioaccumulation factor -for radionuclide "i" in fish (from Table A-1 of Regulatory Guide 1.109, Rev..1), pCi/kg per pCi/1 DF Dose convers i on factor for radi onucl i de "i "

for adults for a particular organ ~ (from Table E-11 of Regulatory Guide 1.109, Rev,. 1),

mrem/pCi I Table 2.2-1 presents the Ai values for an adul t receptor in the SHHPP locale. Values of exp (-x.t 1 p

) are presented in Table 2.2-2 for each radio-nuclide "i." The sum of the cumulative dose from all batch and any continuous releases for a quarter:is compared to one-half the design objectives for total body and any organ. The sum of. the cumulative doses from all releases for a-calendar year is compared to,the design objective doses. The following rela-tionships should hold for the SHHPP to show compliance with Technical Specifi-cation 3.11.1.2.

For the calendar quarter:

D 1.5 mrem total body (2.2-4)

D 5 mrem any organ (2.2-5)

ODCM (SHNPP) 2-18 Rev. 1.0

For the calendar year:

D 3 mrem total body (2.2-6)

D 10 mrem any organ (2.2-7) where:

D Cumulative total dose to any organ or the total body from all releases, mrem:

The quarterly limits given above represent one-half the annual design objec-tive of 10CFR50, Appendix I,Section II.A. If any of- the limits in Expres-sions'.2-4 through 2.2-'7 are exceeded, a special report pursuant to SHNPP Technical Specification 6.9.2 must be filed with the NRC. This report com-plies with Section IV.A of Appendix I, 10CFR50.

2.2.2 Pro 'ection of Doses I

Dose projections for this section. are required at least once per 31 days (monthly) in Technical Specification 4.11.1.3.

The doses will be projected using Equation 2.2-1. When the operational condi-tions for the projected month are to be the same as for the current month, the source-term "inputs into the equation for the projection can be taken directly from the current month's data.. Where possible, credit for expected opera-tional evolutions (i.e., outages, increased power levels, major planned liquid releases, etc.) should be taken in the dose projections. This may be ac-complished by using the source-term data from similar historical. operating experiences where practical. This may also be accomplished by using the projected Percent Power-Reactor Days for the unit as in the following expres-sion:-

D

=

D

'2 i.e., D2=

D P (2.2-8)

ODCM (SHNPP) 2-19 Rev. 1.0

where:

Past month's dose to total body or any organ, mrem Projected month's dose to total body or any

. organ, mrem For past month: (Average X power) x (Reactor

~

days of operation)

P2 For projected month: (Estimated average power) x (Estimated reactor days of operation)

To show compliance with Technical Specification 3. 11. 1.3, the projected month's dose should be compared as in the following:

D < 0.06 mrem for total body (2.2-9) and D < 0.2 mrem for any organ (2.2-10)

If the projections exceed either Expressions 2.2-9 or 2.2-10, then the appro-priate portions of the liquid radwaste treatment system shall be used to reduce releases of radioactivity..

ODCM (SHNPP) 2-20 Rev. 1.0

TABLE 2.1-1 LI(UID EFFLUENT RELEASE TANKS AND PUMPS No. of PUMP CAPACITY ( pm) Eductor Tank Volume Radiation Tank(2) Tanks Process Recirculation Factor (oal.) Effluent Monitor ID SWST 0 2 25,000 . REM-3542 35 I 0 10,000 REM-3541 35 0 25 25,000 REM-3541 TLIIHS 100 0 25 25,000 REM-3540 Reference SHNPP FSAR Tables ll 5 '-1 and 11,2.1-7 SWST: Secondary Waste Sample Tank WECT: Waste Evaporator Condensate Tank WMT: Waste Monitor Tank TLIIHS: Treated Laundry and Hot Shower Tank ODCM (SHNPP) 2-21 Rev. 1.0

Flffult 2.1.2 LEOUIO EF FLUENT FLOW STTEEAM TEIAOTEAM 4

fhEAIED LAUIIIINY1 thEAI ED EJlIPIDNY ~

Naf thaeth tANK Ilaf tllatlth lANK O'E tt htf4 ItfL ttla tfaaNDAht tIAtft SANtLE 'IANK htN-tftftatt t NAfft HEUINALIEAIION tIAIIE MONIIOh WAtft uCWIlah SAEEN tAHK fANK

~ t fILINO tAIIN tlEN ht&tftfLttt I 1

LEOENDg tANK OK tAlIN NAhhl~

LAK t IIAEfI EVAIOIIAIOh COHOEHIAft 'f AllK tfAEIt EVAtahhfah CONDENIAft fAtlK 0 NAOIAIIONtttLUENE tKINffah

FIGURE I 1 3 NORMAL SERVICE WATER F LOW DIAGRAM R LACTOA AUIILIAtt Y I MILO tttC O

C 0

NEAT LOADS EM 5500 WASTE MOCESSINC NEAT LDAO5 5 UILOINC ~

fM 5500 MAINCONDENSER G

Z TVRSINE SUILO INC t

CIRCULATINCWATER UMtS LEGEND REM RADIATIONEltSLUEIIT MONITOR NSW NORMALSERVICE WATER CODLING TOleER SASIN OO C

el

~ .'

el C NOTE: ~ eeteetet Itive ~

Settee Ie etet Seetettt tetet to SSAR et CDOLING TOWEA SLOWDOteN HARRIS LAgE ODCM (SHHPP) 2-27 Rev. 1.0

C) nC7 FIphe 2, I 0 OIIIEIIllOUID EFFI.UENl PA1IIICAYS IUhtlkl CUIlblkO SLOONODALkl tllLUCNILINC IIINCIIitrtba wiltt CVMCC Mhb OIL NlllfhiLIlAIION I LOON ONAINI CCSAhA ION CA SIN

~~

Stlltlkh SASIH OUI SIDC IANKANIAONAIN CIILUIHIUNC OUI Ilhl TANK SIOhM OD AIM AhlAOhAIN Sr l'It M IM

'AN SE DIVChfCP 10 CtCONOAhr WAllt lhtAIMlNIcrsl CM "CANCI PIVINICDIO LIOUIONAOWACIC lhtAlutklCrtftN

"'lhl INILVLNI IOrllh CLOwbOwN COINS ID lht COOLINO Il lhl CAINE LINC NAhhll LAIC INI LULNI tOIIII INOICAICD IN I IOUDC I.I l IP

3, l.i.4 Determine Cm, the maximum accePtable total radioactivity concentra-tion of all noble oas radionuclides in the gaseous effluent t33Ci/cci.

(2. 12 E-3). 0 Cm F ~ f NOTE: . 1lse the 'lower of the O values obtained in Sections 3.1.1.2 and 3.1.1.3. This will protect both the skin and total body from being exposed to the limit.

where:

Use the actual. effluent flow rate or the maximum effluent flow rate at the point of release (cfm) based on design flow rates given below:

22,t350 cfm (Turbine Bldg. Vent Stack 3A).

207,000 cfm (Waste Processing Bldg. Vent Stack 5).

103,500 cfm (Waste Processing Bldg. Vent Stack 5A).

'390,000 cfm (Plant Vent Stack 1). When contain-ment preentry purge occurs, this should include an additional 33,700 cfm.

Release flow rate for batch releases, if applicable (cfm),.

2.12 E-3 = Unit conversion factor to convert uCi/sec/cfm to gCi/cc.

NOT.:: The F values were taken from the FSAR, Chapter 3, Amendment 15, Table 9.4.0-2.

3.1.1.3 Deterfiine CR, the calculated monitor count rate above background attributed to the noble gas radionuclides tcpmj by:

CR QDCM (SHHPP) 3-4 Rev. 1.0

m Obtained from the applicable effluent monitor ef f i ci ency (cpm/uCi/cc) .

3 1.1. 6 Determine the HSP, the moni tor high-alarm setpoint including back-ground fcpm) by:

HSP TmCR + Bkg (3. 1-5) where:

m Fraction of the radioactivity from the site 'that may be released .via the monitored path~ay to en-sure that the exclusion boundary limit is not exceeded due to simultaneous releases from several pathways.

0.03 for Turbine Bldg. Vent Stack 3A.

0.29 for Waste Processing Bldg.'ent Stack 5.

0.14 for Waste Processing Bldg. Vent Stack 5A.

0.54 for Plant Vent Stack l.

Bkg The background count rate (cpm) due to internal contamination and the radiation levels in the area in which the monitor is ins alled when the detec-tor sample chamber is filled with uncontaminated ail ~

Hote: The vent stack monitors are designed such that the high-alarm setpoint can be input. as uCi/sec or uCi/cc. The monitor setpoint in uCi/sec can be obtained by multi-plying the lowest q value (obtained from Sections ODCM (SHNPP) 3-5 Rev. 1.0

3.1.1.2 and 3.1.1.3) by the T value found in Section

3. 1. 1.6. The uCi/cc setpoint can be obtained by dividing the uCi/sec setpoint by the design or process flow rate in cc/sec. The equations for calculating the setpoint in cpm are included for completeness and may be used if desired.

3.1.2 Alternative Setooint Determination Method Based on Gaseous Effluent Analysis Prior to Release The following method applies to setpoint determinations. from plant vent stacks during the operational conditions listed below and when the gaseous effluent

's sampled prior to release:

~ Batch mode release of containment pressure relief.

Batch release of waste gas decay tanks.

3.1.2.1 Determine the maximum allowable discharge flow rate prior to dilu-tion.

. a. Determine f, 'the maximum acceptable, gaseous flow rate from con-tainment or from the waste gas decay tanks (cfm), based upon the whole body exposure limit by:

0.848 T where:

Fraction of the radioactivity from the site that may be released via the monitored pathway to ensure that the exclusion boundary limit is not exceeded due to simultaneous releases from several pathways (see Section 3.1.1.6 earlier).

ODCM (SHHPP) 3-6 Rev. 1.0

5.09 A combined conversion factor consisting of the skin dose limit of 3000 mrem/yr, times a conversion. constant of 2. 12 E-3 to convert cc/sec to cfm, times 0.80, an engineering factor to prevent spurious alarms.

c. The rate at which the noble gas, activity is released from the containment during purging or pressure relief or from the waste gas decay tanks shall not exceed the smaller of the two "f" val-ues calculated in Steps..a and b above.

.3.1.2.2 Determine the monitor setpoint equivalent to the maximum allowable discharge flow rate:

Determine Cm, Che maximum -radi oacti vi ty concentration of al 1 noble gas radionuclides to be released during containment purge or pressure relief via Plant Vent Stack 1 or waste gas decay tanks discharge via the Waste Processing Bldg Vent Stack 5 after by other discharges in the respective stacks (uCi/cc): 'ilution I

C F+f where:

Ct The total radioactivity concentration of all noble gas radionuclides in the gas to be discharged from the containment or waste gas decay tanks prior to dilution (uCi/cc).

acceptable gaseous -flow rate 'rom The maximum containment 'r -from the waste gas decay tanks (cfm) .

The maximum design vent stack flow rate (see Section 3.1.1.4 earlier or the actual flow rate). 3E.

I ODCH (SHHPP) 3-8 Rev. 1.0

Determine CR, the calculated monitor count rate above background attributed to tne radionuclides [cpm).

CR is obtained by using the applicable effluent monitor effic iency "Em" (cpm/qCi/cc):

CR {Cm) (Em) (3.1-9)

c. Determine HSP, the monitor high-alarm setpoint including back-"

[cpm] by: 'round HSP CR + Bkg (3.1-10) where:

Bkg Monitor background (cpm)

I d.. The monitor HSP shall be set at or below the calculated. value during containment purges or, releases from the waste gas decay tanks. If containment pur ges or pressure re 1 i ef or waste gas decay tanks releases are made while other sources of noble gas activity are being released from their respective stacks, the monitor HSP shall not exceed the calculated value determined in Section 3.1.1.

3.1.3 Alternative Setooint Determination Based on Gaseous Effluent'Analysis Prior to Release and Estimates of Maximum Acceptable Flow Rate The following method applies to gaseous releases when the maximum acceptable effluent flow rate at the point of release is given and the associated high-alarm setpoint based on this flow rate is de-sired. The method is applicable during the following operational conditions:

ODCM (SHHPP) 3-9 Rev. 1.0

~ Batch release of containment purge via Plant Vent Stack l.

Batch release of containment pressure relief via Plan'. Vent Stack 1.

Batch release of waste gas decay tanks via Waste Processing Building Vent Stack 5.

3.1.3.1 Determine G;, the noble gas release rate for radionuclide "i," ))Ci/sec Gi 472 (C'i (F (3. 1-11.)

where:

472 = 472 cc/sec/cfm Ci The radioactivity concentration of noble gas radio-nuclide "i" in the gaseous effluent from the analysis of the gaseous effluent to be released, )2Ci/cc F = The maximum acceptable effluent flow rate at the point of relea.se, cfm

.30 for one condenser vacuum pump 33,700 for one containment purge pump 2.26 66 (

14.7 ) (

2730 T

)

for containment pressure relief t

ODCH (SHHPP) 3-10 Rev. 1.0

t coo( 14.7 ) {

273o T

t

)

for a waste gas decay tank release where:

2.26 E6 and 600 are the volumes in ft3 of the containment and decay tank, respectively, and T , Tt, n Pc, and A Pt are the estimated, respective temperature and change in pressure (psig) following the release of the containment and decay tank; and, 14.7 = lb/in2, i.e., 1 atmosphere pressure Length of release, min 273'K - 0 C Tt'c 273 K + C 3.1.3.2 Determine the monitor alarm setpoint based on total body dose rate:

a. Determine'Q (the monitor count rate per mrem/yr, total body)

C CR (3.1-12)

(Xlq) z. K.G.

where:

C = The count rate of the monitor corresponding. to the radioactivity concentration in the analyzed sample (C [Ci] )the monitor efficiency])

X/g The highest calculated annual average relative disper-sion factor for any area at or beyond the exclusion boundary for all sectors (sec/m 3 ) from Appendix A.

ODCH (SHNPP) 3-11 Rev. 1.0

2.06 E-6 sec/m 3 from Table A-1, Appendix A V,.

1

= The total whole .body dose factor due to gamma emissions from noble gas radionuclide "i" mrem/yr/~Ci/m from Table 3.1-2

b. Determine St, the count rate of the gaseous effluent noble gas monitor at the alarm setpoint based on total body dose rate,, cpm:

S = ISF T ~ D ~ CR I + Bkg (3.1-13) t m t t

'here:

SF An engineering factor used to provide a margin of safety for cumulative uncertainties of measurements.

- 0.5 Dt.' .500 mrem/yr, the total body dose rate 'limit Tm Fraction of the radioactivity from the site that may be released via the monitored pathway to ensure that the exclusion boundary limit is not exceeded due to simul-taneous releases from several pathways (see Section 3.1.1.6 earl ier)

Bkg = The background count rate due to internal contamination and the radiation levels in the area in which the moni-tor is installed when the detector sample chamber is f'illed with uncontaminated air, cpm 3.1.3.3 Determine the monitor alarm setpoint based on the skin dose rate:

a. Determine CRs (the monitor count rate per mrem/yr, skin):

ODCM (SHHPP), 3-12 Rev. 1.0

CRs where:

x/Q i

z. (L.

i

+ 1.1 M.) (G.)

, i i (3.1-14)

+ 1.1 Ni The total skin dose factor due to emissions from

'oble gas . radionuclde "i" (mrem/yr/uCi/m 3 ) from Table 3.1-2

b. Determine S, the count rate of the gaseous effluent noble gas monitor at the alarm setpoint based on the dose rate to the skin, cpm S = [SF - T D ~ CR' + Bkg (3. 1-15) s m s s where:

Bkg = 'The background count rate due to internal contamination and the radiation levels in the area in which the moni-tor is installed when the detector sample chamber is f'illed with uncontaminated air, cpm

.Ds ,. = 3000 mrem/yr, the dose rate to the skin limit 3.1.3.4 Determine the actual gaseous monitor setpoint:

The respective monitor setpoints, based on the dose rate limits to the tota1 body (St) and to the skin (Ss), are compared and the lesser value is 'used as -the monitor HSP; i.e., high-alarm setpoint. If containment purges or pressure re1ief or'aste gas decay tanks re-leases are made while other sources of noble gas activity are being released from their respective stacks, the monitor HSP sha'11 not exceed the calculated value determined in Section 3.1.1

3. 1.4 Effluent Honitorina During Hoooino Operations ODCM (SHHPP) 3-13 Rev. 1.0

If the reactor has been shut down for less than 30 days, the conden-ser vacuum discharge during initial hogging operations at plant start-up and prior to turbine operation will be routed directly to Turbine Building Vent Stack 3a. In this event, the setpoint methodo-logies of Sections 3. 1.1 and 3.1.2 for the noble gas monitor located on Vent Stack 3a (see Appendix D) are applicable.

the reactor has been shut down for greater than 30 days, the condenser vacuum pump discharge during initial hogging operations at plant start-up and prior to turbine operation may be routed as dual exhaust to (1) the Turbine Vent Stack 3a and (2). the atmosphere directly. En this instance, the blind flange on the latter exhaust route will be removed (see Figure 3.3).

Setpoint determination in this case depends on knowledge of the flow rates through each of the exhaust pathways. Once these flows are established or estimated, the ratio of the flow through Vent Stack 3a 0

to the flow in the direct exhaust path will be computed. This ratio will be used to reduce the setpoint on Vent Stack 3a to account for noble gases being exhausted concurrently via,dual pathways.

ODC~ (SH~pp) 3-14 Rev. 1.0

. TABLE 3.1-1 GASEOUS SOURCE TERHS*

Condenser Air Containment Purge Plant Vent Release via Vacuum via or Presure Relief via Gas Decay Tanks via Vent Stack 1

" Vent Stack 3A

. ~

Vent Stack 1 Vent Stack 5 Rad l onuc 1 i de A l (C l /yr) S i Al (Cl/yr) Al (Ci/yr) Sl Al (Ci/yr) S l

Kr-83m O.OOE 00 O.OOE 00 O.OOE 00 O.OOE 00 1.0E 00 3,78E-04 O.OOE 00 O.OOE 00 Kr-85m 3.0E 00 2. 16E-02 2.0E 00 2.44E-02 1.2E 01 4.53E-03 O.OOE 00 0.00E 00 Kr-85 O.OOE 00 O.OOE 00 O.OOE 00 O.OOE 00 . 4.0E 00 1.51E-03 2.1E 02 9.81E-01 Kr-87 1.QE 00 7. 19E-03 O.OOE 00 O.OOE 00 Z.OE 00 7.56E-04 O.OOE 00 O.OOE 00 Kr-88 5.5% OO 3.60E-02 3.0E 00 3.66E-02 1.6E 01 6.05E-03 O.OOE 00 O.OOE 00 Kr-89 O.OOE 00 - O.OOE 00 O.OOE 00 O.OOE 00 0.00E 00 O.OOE 00 O.OOE 00 O.OOE 00 Xe-131m O.OOE 00 O.OOE 00 O.OOE 00 O.OOE 00 1.0E 01 3.78E-03 3.0E 00 1:40E-02 Xe-133m 2.0E 00 1.44E-02 ).OE 00 1.22E-02 4.3E 01 1.62E-02 '.00E 00 0.00E 00 XB-133 1.2E 02 8.63E-01 7.2E 01 8.78E-01 2.5E 03 9.44E-01 1.0E 00 4.67E-03 Xe-135m 0.00E 00 O.OOE 00 O.OOE 00 O.OOE 00 O.OE 00 O.OOE-01 O.OOE 00 0.00E '00 Xe-135 7.0E 00 5.04E-02 4.0E 00 4.88E-02 5.9E 01 2.23E-02 0.00E 00 O.OOE 00 Xe-137 O.OOE 00 O.OOE 00 O.OOE 00 O.OOE 00 O.OOE 00 O.OOE 01 O.OOE 00 O.OOE 00 XQ-138 1.OE OO 7. 19E-03 O.OOE OO O.OOE 00 O.OOE 00 O.OOE 01 O.OOE 00 O.OOE 00 TOTAL 1.39E 02 8.20E 01 2.64E 03 2.14E 02 Source terms are based upon GALE Code (see SilHPP FSAR Table 11.3.3-1) and not actual releases. These values only apply to routine releases and should not be taken as a complete inventory of noble gases ln an emergency s l tuat ion.

Li The skin dose factor due to beta emissions for noble gas radionuclide "i," mrem/year per gCi/m .

The air dose factor due to gamma emissions for noble gas radionuclide "i," mrad/year per gCi/m .

The ratio of the tissue to air absorption coeffi-cients over the energy range of the photon of interest, mrem/mrad (Reference NUREG-0133).

The release, rate of noble gas radionuclide "i" in gaseous ef fluents f rom al l plant vent stacks .

(uCi/sec).

The determination of limiting location for implementation of 10CFR20 for noble gases is a function of the radionuclide mix, isotopic release rate, and the meteorology.

The radionuclide mix was based upon source terms calculated using the NRC GALE Code and presented in the SHNPP FSAR Table 11.3.3-1. They are reproduced in Table 3.2-1 as a function of release point.

The X/g values utilized in the equations for implementation of 10CFRZO are based upon the maximum long-term annual average (X/g) in the unrestricted area. Long-term annual average {X/Q) values for the SHNPP release points to the special locations in Table 3.2-2 are presented in Appendix A. A descrip-tion of their derivation is also provided in this appendix.

To select the limiting location, the highest annual average X/g value for ground-level releases is the'ontrolling factor. Long-term annual average {X/g) values were calculated 'assuming no decay, undepleted transport to the exclusion. boundary, and are given in, Table A-l, Append ndixx A . The maximum exclusion boundary X/g for ground-level releases occurs at the NNE and SSW sectors. However, the limiting location for implementation of 10CFR20 for noble gases is considered to be the exclusion boundary (1.33 miles) in the NNE sector due to the generally greater population density in this direction.

OOCM (SHNPP) 3-1S Rev. 1.0

up n the source terms calculated using the GALE Code.

s d upon again b ase The mix and erms are presented in Table.3.2-1 as a function of release point.

the source terms t

The ddetermina'ion ion oof the controlling exclusion boundary location was based upon ththe h ig hest es eexclusion boundary 0/g value. The determination of actual t 1 hamiiting receptor ing location o was based upon the milk pathway 0/g value and the P'alue for the respective milk path~ay; Values for P; were calculated for an infant for various radionuclides for the .inhalation, ground plane, cow milk, and goat milk pathways using the, methodology of HUREG-0133. The P; values are presented in Table 3.2-4. A description of the methodology used in calculating the Pi values is presented in Appendix B. The values of P; re-flect, for each radionuclide, the maximum P; value for any organ for .each individual pathway of exposure. The goat milk pathway is present near SHNPP, as is the cow milk pathway. \

However, the cow milk pathway Pi values were utilized in the determination of the controlling location because the product of the maximum cow milk pathway 0/g and P-1 values were greater than those for the goat. For the case of an infant being present at the site at the exclusion boundary.-or :at the real path~ay location, the ground plane pathway is not considered as a reasonable exposure pathway (i.e., Pi 0). However, P; values are presented in Table lG 3.2-4 for completeness.

The annual average [0/qJ values at the special locations, which will be uti-lized in Equation 3.2-3, are obtained from the tables .presented in Appen-dix A. The [X/g] values which will be utilized in Equation 3.2-3 are also obtained from the tables presented in Appendix A. A description of the deri-vation'f the X/g and 0/g values is provided in Appendix A.

ODCM {SHNPP) 3-21 Rev. 1.0

power levels, major planned liquid releases, etc.) should be:aken in the dose projections. This may be accomplished by using source-term data from similar.

historical operating experiences where practical. This may also be .accom-

. plishe using the projected percent power-Reactor Days for the unit as in h d b y usin the following expression.

'g i.e- 0 2 P (3.3-7)

P1, P 1

where:

Past month's dose to total body or any organ, .mrem D2 Projected month's dose to total body or any organ, mrem PI For past month: (Average " power) x (Reactor days of operation)

P2 For projected .month: (Estimated average ~ power) x (Estimated reactor days'of operation)

To show .compl iance with Technical Specification 3.11.2.4, the "projected month's dose should be compared as in the following:

D < 0.2 mrad to air fo'r gamma radiation (3.3-8)

Y D

B

< 0.4 mrad to air for beta radiation (3.3-9)

If the projections exceed either Expressions,3.3-8 or 3.3-9, then the appro-Ea priate'ortions of the 'aseous radwaste treatment system shall be used to reduce releases of radioactivity.

DOCH (SHHPP} 3-29 Rev. 1.0

R Dose factor for an organ for,radionuclide "i" for I

the inhalation pathway, mrem/yr per uCl/m 3 .

~

Ri Dose factor for an organ for radionuclide "i" for V

the vegetable pathway, mrem/yr per uCi/sec per m

Ri Dose factor for an organ for radionuclide "i" for 8 -2 the meat pathway, mrem/yr per uCi/sec per m

~

Dose factor for an organ for tritium for the milk

~

3 pathway mrem/yr per uCi/m .

R Ty Dose factor for an organ for tritium for the vege-3 table pathway, mrem/yr per uCi/m .

RT Dose factor for an organ for tritium for the inha-I lation pathway, mrem/yr per uCi/m 3 .,

~

RT Dose factor for an organ -'for tritium for the meat 8 ~ 3 pathway, mrem/yr per uC>/m .

<Tv Release of tritium in. gaseous effluents for long-term vent stack releases (> 500 hrs/yr), pCi.

qTv Release of tritium in gaseous effluents for short-term vent stack releases (< 500 hrs/yr), uCi.

To show compliance with .10CFR50, Equation 3.3-10 is evaluated at the limiting real pathway location. At SHHPP this location is 2.2 miles in the H sector.

The critical receptor is an infant. Appropriate X/g and D/g values from tables in Appendix A are used. For this document,'Song-term -annua'1 average X/9 and D/g values may be used in lieu of short-term values (see Section 3.0 earlier).

ODCM (SHHPP) 3-32 Rev. 1.0

0, operational conditions for the projected month are expected to be the same as for the current month, the source-term inputs into the equation for the pro-jection can be taken directly from the current month's data. Where possible, credit for expected operational evolutions ('i.e., outages, increased power levels, major planned liquid releases, etc.) should be taken in the dose projections. This may be accomplished by using source term data from similar historical operating experiences where practical. This may also be accom-plished by the using projected Percent Power-Reactor Days for the unit as in the following expression:

D2 1.e.,' 0 {3.3->3)

P2 PI where:

Past month's dose to total body or any organ, mrem Projected month's dose to total body or any organ, mrem c For past month: (Average " power) x (Reactor days of P1 operation)

V P2 For projected month: (Estimated average " power) x (Estimated .reactor days of operation)

To show compl i ance with Techni cal Speci fication 3.11.2.4, the pro jected month's dose should be compared as in the following:

D < 0.3 mrem to any organ (3.3-14)

If the projections exceed Expression 3.3-14, then the appropriate portions of the gaseous radwaste treatment system. shall be used to reduce releases of radioactivity.

ODCM (SHNPP) 3-35 Rev. 1.0

Flgur ~ 3.1 SIINPP GASEOUS WASTE STllEAMS UNIT T ntll ~ HADIATION lt ILUtHT MONITon WIS ~ WASTE tn4CElllNO SLD4 HAS ~ HEACToh AUKILIAHY~ LD4 IH'I f UIIEINE bLITO VENT SlACK 1A SsK II&IT1I ~ Is ~ Irrrrf 'NAOH r FUEL HAHOLIHO ~ LDO r

gg rxr Ifv' ll~

CONDSNSth VACWMtIXKF WAETE TIIQCEEE INa aLOa VEN T a TACK b Wtf HOT b COLO LAVNDhl 1~ r Aslo Irv ~ I ~ ~

. ~

wts oftlct AntA ExHAUIT wte colo LAUNonY Dnf tnt WFS OFFICE AHEA wts coNlnoL noou suoxt EXHAUTT Wtl CHILLEH noolf EXHAUlf Wtl OtHEHAL ADEA EXHAUlf WASTE tnoCESIIN4 AhlAS FILI tnt D fXHAUlf WASTE OAS DECAY TANKS TTAETE FIIOCETIINO ~ LM VENT ETACK EA rslrv I ~~ II lrr ~ ssl Ak 1( rx&lrrl ~~ I Wtf SWITCHOE Ahhoofl f XHAUlf wts HYAc EDUlt. nooQ E xHAvs I wts ttnsoHHELHAHDLIHOFAcILI'IYExHAvs'I Wtl llof ~ LOWACTIYITYEXIIAUST litt LAS AHEA EXIIAUlf PLANT VENT StACK I XIIIAv'I I>l

~ ~A rx ~ rl xr al 1xlalxr I ~ Itx

~

XIH CONf AhxltNf thE ENThY FUnof HAS HOOIIAL EXHAUST III'AYI II ~ FHS HohllAL EXHAUST NOHTH IIKRIAY~ I ~ ~ A appal FHS HonllAL txHAVST SOUTH nAS f VlnolHCY EXHAUST ss ~

Axial~ Isl ~ ~ ~ FHS HOnllALtXHAVSI lot tn. Tl..f SOUTH HAS VfHIILATION SY Sf ELI KK FHS HonllAL EXHAUST lot th. TLI SOUTH HYDDOOEN tlxlot axr AKMItlsl~ K~ A Oa

'AS SMOKE FIXIOE K~ M FHS EME notNCY EKHAVSf rxwIKDIK~ ~ ~ ~ ~

nxs tvnGE 1 ~ Ks tl'lssss ~ ~ l~ ~ I I~ ~ I OMrxx ~ roL

flW5f IIEFVTLfNOWATth STOhAOE TANK hMWST htACTORMAKEVCWAl'thSfORAGETANK CRAr r CARRIna CST COHOENSATE SIOhAOE TANK I LANT NORTII MAONE'TIC HORTII VEIIT STACK J

0 E

r I

J W

IN IT F28 TX'CESSI VENT STACK P

CST COOLINO TOWER nwsr gw EATMEI Pi3 SERVICE 'ECV 1Lll4 FARRINO AREA 0 OO WA ll IOV E SWI f CNVARO RARIIINO AREA t

5IIEAflONIIARhfS NUCLEAR COWT fl CLANI CAflOLINAFOWEh S LICIIT COMI'ANY SCIIEMATIC OF >LAHf AlflSOtlNE EFFI.VENT hELEASE FOIHTS F IOVflE 2.2

Flpure 3.3 SllNPP CONDENSEA OFF GAS SYSTEM 0

STACKS.-

TUREIHE ELDO litt IfV till IHtlOMI VENT SA GLAIIO Sf EAM COND.

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gb r

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FIGURE 4.l4 SHEARON HARRIS NUCLEAR POWER PLANT

'I

\ ~

i ~ ENVIRONMENTALRADIOLOGICALSAMPLING POINTS

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Ma "'

/t II gII 'Il

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FIGURE 4.1<

ERIENOSHII'HEARON HARRIS NUCLEAR POWER PLANT ENyIRPNMENTALRADIOLOGICALSAMPLING POINTS N a NEYI HILL I

'l I INg BONSAI.

Or MERRY OAKS hg a

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~EXCLUSION SOUNOARY 00o I.

I.

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4-aS

FIGURE 4.1 5 LEGEND STATION STATION SYMSOL NuMddh 5YMSOL NUMddII AP, AC. TL 0 ~ AP.*C. SW. SS.'TL o

Al, AC TL 0$ ' TL 0

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TL 42 15 TL MC, SC 15 MrLTL CH

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7.0 LICEHSEE-IHITIATED CHAHGES TO THE ODCM Pursuant to Technical Specification 6.14.2, licenseo-initiated changes to the Off-Site Dose Calculation Manual:

A. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change(s) was made effective. This submittal shall contain:

1. Sufficiently detailed information to totally support the rationale f'r the change without benefit of additional or

.supplemental information. Information submitted should consist of a package of those pages of the ODCM changed with each page numbered, dated, and containing the revision number together with appropriate analyses of evaluations'justifying the change(s). .

2. A determination that the change will not reduce the accuracy or reliability= of dose calculations or setpoint determinations,

3. Documentation of the fact that the change has been reviewed and found acceptable by-the PHSC.

B. Shall become effective upon review and acceptance by the PHSC.

ODCM (SHHPP) 7-1 Rev. 1.0

APPENDIX D RADIOACTIVE LIQUID AHD GASEOUS EFFLUEHT MONITORING IHSTRUMEHTATIOH NUMBERS Monitor Li uid Effluent Monitorina Instruments Identification

>> A. Treated Laundry and Hot Shower Tank............. ....... REM-3540 B Waste Monitor Tank.. .'.................................. REM-3541 C. Waste Evaporator Condensate Tank........................ REM-3541 D. Secondary Waste Sample Tank..................;..... REM-3542 E. Hormal Service Water Returns

--- to Circulating Mater System From Waste Processing Building............. ...... REM-1SW-3500A .

From .Reactor Auxiliary Building...................... REM-1$ W-3500B F. Outdoor Tank Area Drain Transfer Pump Monitor........... REM-3530 G. Turbine Building Floor Drains Effluent.............;.... REM-3528 Gaseous Effluent Monitorina Instruments A. Plant Vent Stack 'l.

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'1.. Plant Vent Stack '1................................. REM-lAV-3509-SA RN-1AV-3509-1SA>>

.'.Reactor Auxiliary Building Normal Exhaust............ REM-1AV-3531

'...3. Reactor Auxiliary Building Emergency Exhaust........ REM-1A-3532A 4 'uel Handling Building Normal Exhaust (South)....... REN 1FL-3506

5. Fuel Handling Building Normal Exhaust (South)....... REN-1FL-3507

6. Fuel Handling Building Emergency Exhaust............ REN-1FL-350BA-SA

7. Fuel Handling Building Emergency Exhaust........... REN-1FL-350BB-SB Containment Pre-Entry Purge.............. .'.... ....

'. B.

Turbine Building Vent Stack 3A........... .... .........

REN-1LT-3502B RM-1TV-3536-1>>

1. Condenser Vacuum Effluent Line..... REM-1TV-3534 C. Waste Processing Building Vent Stack 5.......... ....... REM-1WV-3546 RM-1WV-3546-1>>

D. Waste Process'ing Building Vent Stack 5A.. .............. REN-1WV-3547 RM-1WV-3547-1

• Wide-Range Gas Monitor (WRGM)

ODCM (SHNPP) D-l Rev. 1.0

Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 5 : Changes to the Environmental Monitoring Program Enclosure 1 : Environmental Monitoring Program Technical Specifications 3.11.2.3

3. 12. 1 3.12.1.c No changes have been made to the Environmental Monitoring Program during this report period.

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Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 5  : Changes to the Environmental Monitoring Program Enclosure 2 : Land Use Census Technical Specifications 3.12.2.a 3.12.2.b A land-use census was performed in May of 1987. Comparison with the 1986 land-use survey indicates the following changes:

A. Milk goats were not located within the five-mile radius.

B. Milk cows are presently located in the N and NNE sectors. These locations are commercial dairies that are currently included in the SHNPP environmental sampling program.

Table 1 summarizes the location of the nearest milk animal, meat animal, residence and garden in each of the 16 compass sectors.

Table 2 lists the kinds of meat animals at each meat animal location. Cattle and hogs are the predominate animals nearest the site.

TABLE 1 DISTANCE TO THB NEAREST SPECIAL LOCATIONS FOR THE HARRIS NUCLEAR PROJECT (MILES)

EXCLUSION I. MILK MBAT SECTOR BOUNDARY RESIDENCE ANIMAL GARDEN ANIMAL N 1.32 2.2 2.2 2.2 2.2 NNB 1.33 1.7 '.6 1.7 1.8 NE 1.33 2.3 2.3 2.3 BNE 1.33 2.0 2.0 E 1.33 1.9 4.7 4.6 ESB 1.33 2.7 2.8 4.4 SE 1.33 4.7 4.7 SSE 1.36 4,4 S 1.33 SSW 1.33 3.9 3.9 4.4 SN 1.33 2.8 2.8 2.8 NSW 1.33 4.3 4.3 4.3 W- 1.33 2.7 3.0 3.1 lQM 1.33 2.1 2.1 2.5 NN 1.26 1.8 3.8 3.8 1.26 1.5 1.9 1.9 5/2

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Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 5  : Changes to the Environmental Monitoring Program Enclosure 2 : Land Use Census Technical Specifications 3.12.2.a 3.12.2.b TABLE 2 MEAT ANIMAL TYPE AT NEAREST LOCATION TO SHNPP BY SECTOR DISTANCE MEAT SECTOR (MILES) TYPE OWNER N 2.2 HOGS GOODWIN NNE 1.8 BEEF GUNTER NE 2.3 BEEF / GOATS JAMES REST HOME ENE 2.0 RABBIT / FOWL HARRIS E 4.6 BEEF McIVERS ESE 4,4 HOGS PATTERSON SE SSE S

SSW 4,4 HOGS CROSS SW 2.8 4.3 BEEF / FOWL / GOATS POLLARD WSW HOGS SMITH, P.

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Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 6 : Additional Technical Specification Responsibilities Enclosure 1 : Inoperability of Liquid Effluent Monitors Technical Specification 3.3.3.10, Action b Monitors out-of-service ) 30 Days for the Period After Receipt of Operating License (10/24/87) and Befoxe January 1, 1987 Radiation Days Monitor Inop. Reason REM-01MD-3528 .62 Modification required to ensure monitoring Turbine Building of effluent stream when sump pump actuates.

Drains REM-21WL-3541 39 Monitor does not correspond with analyzed Waste Monitor results due to high sample chamber background.

Tank Investigation and procedural changes needed for determining new monitor setpoints and background subtract values.

REM-1SW-3500A 39 Monitor in Pre-op testing.

WPB Normal Service Water Monitor REM-1MD-3530 39 Monitor in Pre-op testing.

Tank Area Drains REM-1SW-3500B 54 Modification required to relocate sample line.

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Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 6  : Additional Technical Specification Responsibilities Enclosure 1 : Inoperability of Liquid Effluent Monitors Technical Specification 3.3.3.10, Action b Monitors out-of-service > 30 Days For the Report Period Radiation Days Monitor Inop. Reason REM-01MD-3528 181 Modification required to ensure monitoring Turbine Building of effluent stream when sump pump actuates.

Drains REM-1WL-3540 71 Monitor does not correspond with analyzed Treated Laundry results due to high sample chamber background.

and Hot Shower Investigation and procedural changes needed for determining new monitor setpoints and background subtract values.

38 Same as above REM-21WL-3541 59 Monitor does not correspond with analyzed Waste Monitor results due to high sample chamber background.

Tank Investigation and procedural changes needed for determining new monitor setpoints and background subtract values.

Same as above.

REM-1WS-3542 Monitor detector damaged by high temperature Secondary Waste water. Modification required to provide Sampling Tank cooling water for sample line.

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II (f h, WIH}t ~ i I <<>> I W hl'II 'l'>>'I I Il W>> Il II,, t}'< ,,>>W \'I W ~ <<lf< ">> W ' t I'h ~ ( h W H ',>>t WW ","'JW H,I W""C g"j <<I<< II Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 6  : Additional Technical Specification Responsibilities Enclosure 2 : Inoperability of Gaseous Effluent Monitors Technical Specification 3.3.3.11, Action a Monitors out-of-service > 30 Days For the Report Period Radiation Days Monitor Inop. Reason Turbine Building 75 Flow monitoring problems due to excessive Stack 3A moisture in the sample lines. Modification Flow Rate Monitor required to install moisture control unit. WPB Vent Problems with calibration of flow control Stack 5 system resulting in discrepancies between Flow Rate Monitor actual and expected flow rates. 119 Same as above. %'B Vent 181 Problems with calibration of flow control Stack 5A system resulting in discrepancies between Flow Rate Monitor actual and expected flow rates. 66 Same as above. Iform i~ii ) ~) i ~ t I','fW .( i((II" '('t)'S )) )<R<, If (II', " Ijl'if I, II)("( r ~ I'[ ) 1+i I 2 fI' ~ If,<') <<)f'( ,i)f II< '(( ~ . I ASCII 1I,'" (* l Eg )a~.'f I E> ) i)~lIi i" Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 6  : Additional Technical Specification Responsibilities Enclosure 3 : Unprotected Outdoor Tanks Exceeding Limits Technical Specification 3.11.1.4, Action a No unprotected outdoor tank exceeded the Technical Specification limit during this reporting period. 6/5 ~ fl lI )' II ", ll I l' ' 'l k ~'t I I I, I ll Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 6 : Additional Technical Specification Responsibilities Enclosure 4 : Gas Storage Tanks Exceeding Limits Technical Specification 3.11.2.6, Action a No gas storage tank exceed the Technical Specification limit during this reporting period. 6/6 I P 4 I 4I If 44 ~ 'I 4I 4II4 4I I pl[1 W II, 4 4 f,,, II 44 v Semiannual Radioactive Effluent Release Report January 1, 1987 to June 30, 1987 Appendix 7  : Major Modifications to Radwaste System Technical Specification 6.15.1 RADWASTE SOLIDIFICATION SYSTEM Functional Summer : The original design of, the Radwaste Solidification System did not provide the capability to hook up a vendor's mobile solidification system as a backup to the installed solidification system. This modification allows a vendor to hook up a mobile unit to the plant Solidification System Pretreatment Tank, Spent Resin Storage Tank, and Filter Particulate Concentrates Tank. Mobile solidification and resin dewatering services were installed in March 1987. Safety Summary: The modification was reviewed in accordance with 10 CFR 50.59 and found not to be an unreviewed safety question. The consequences of a spill of the Liquid Waste Processing System will not increase since the contents will be contained within the Waste Processing Building (WPB). The modification will not increase the inventories or sources contained in the WPB which hav'e already been analyzed. Reason for Change: Change was required due to'tartup Testing of the Radwaste Solidification System was not completed and the system was not operable. In accordance with Technical Specification 3.11.3 contract capabilities must be available when the installed Solid Radwaste System is not operable. == Description:== Three spool pieces were added to the inlet lines to the installed solidification system. These spool pieces allow waste to be routed to the plant solidification system or connected to lines which direct the waste to the future drum storage area located adjacent to the truck loading bay on level 261'f the Waste Processing Building. These spool pieces allow waste evaporator bottoms and chemical waste from Solidification Pretreatment Tank B, spent resin and filter particulates concentrates to be sent to a mobile solidification system and provides positive isolation between the plant solidification system and the vendor system. Service connections for service air, demineralized water, and a connection to the Waste Processing Building floor drain system is provided. Penetrations for lines from the bulk vendor chemical trailer are provided through the east wall of the truck loading bay. Vendor solidification and resin dewatering services were contracted in March 1987. These services are provided by Chem-Nuclear Systems, Inc. and are described in Topical Reports, CNSI-2-4313-01354-01-A, Mobile Cement Solidifi-cation System and CNSI-DW-11118-01-NP-A, Dewatering Control Process Contain-ers. 7/1 1, I'" "' >>I (tf ft N "(NK h ) f<<, I'l/ IN FN I,, 'tl<< ft tlf ft, t .I ~H *1'4 1 lf ff'l . '>>)', 1 It,f(<<,<<, ' lf" 1 1 4 t t 4 (~ If 5 HW I '1 9'l (1 4'(f 1 IH t,', WPJ, 1, <<f lfffh( 1, KFK Il 1 4'Uh 1 f, I'F(fl h ~ 'N<<lt 1 hl g'1 ~ K 1 4 1'ft 4 'll .' K Nr. 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" ~ J << I 1 As provided in Section 4.2 of the Shearon Harris Nuclear Power Plant Process Control Program, PLP-300, the vendors Process Control Program, CNSI-SDWP-003, is being used to establish processing conditions assuring safe and effective solidification of waste. 10 CFR 61 Waste Form Certification Testing has been completed by Chem-Nuclear System, Inc. and is contained in Topical Report CNSI-WF-Ol-NP. Solidification and dewatering is being performed under the direction and supervision of a Radwaste Shiit Foreman by the vendor's trained operator using vendor's approved procedures. ~uuantit of Solid Waste: Based on the solid waste processing system inputs given in the FSAR Table 11.4.1-1, the projected quantity of solid waste that will be generated using the vendor's service is as follows: Quantity Quantity 195 cu.ft. 100 cu.ft. Source Form cu.ft.gyr liners boxes Spent Resin Dewatered 1,840 (1) 9 Evaporator Bottoms Solidified 10,894 (2) 53 Filter Particulates Solidified 2,733 (2) 13 Dry Solids Compressed 2',000 (3) 20 Chemical Drains Solidified 190 Q2} "17,567 20 Notes: (Bases for values) (1) Based on 180 cubic feet of resin in a 195 cubic foot liner with a burial volume of 205 cubic feet. High integrity containers (HIC) may be used as required. (2) Based on 135 cubic feet of waste in a 195 cubic foot liner with a burial volume of 205 cubic feet. (3) Based on a 6 to 1 volume reduction using a vendor's super compactor ser vice. Exposure to a Member of the Public: No exposure to a member of the public in an unrestricted area different from those previously estimated in the License application is expected from use of the vendor's solidification/dewatering service. Ex~ected~Actual Waste Generated: During the period of March 1987 through June 1987 expected and actual waste generated is given below. Prior to March 1987 no solid waste was generated. Expected(l) Generated Source Form cu.ft. cu. ft. Spent Resin Dewatered 460 205 Evaporator Bottoms Solidified 2,701 1,845 Filter Particulates Solidified 683 0 Chemical Drains Solidified 48 0 3, 892 2,050 7/2 ~ << ><<f)l >>> lf It ff>> r>> I W>') ~ H.f>>>> ri', I >>)M H >> >li r'<>r'fl <<4 I* I f, I>><< "I H 'll f 1' H [ I'>il<< 'I 'I H >>> 1 ! H' I 4 H 'HH <<>IK Uf f ~ H>r l H>>f j)t ~ f r<<. I 'Hf( f.j,,>If,f;,,~,>>,,<<W 4 >., I; f )>H f>>r<<>> I I ir>>II' '<<ff =, r I 'r g<<f';H ff I.f<<>>f ('t ' 'f>>,l r>> 'r

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'E r>> ~ >>>fi ' "'f II f I!Ill >'I, <<) f>> ~ r> H, I I II I'<< , ~ I'<<W S " ><< t "rf <<<< ig lf<< - 'if! r >!<< H. ~ "' H '<<HI'>> fr r 'r., ,r f )<<r' I'I ,)f,)>> j P, <<>>W 'f > PW II f ff I( VII 4<<- <<lf'H <<'<<Iaf><<<< 't f f >> I Hf I 'I ff >~ Hff I>>>> I>> I "'l"f 4 p 'I <<)<<I, f>>>> ffl<< 4 ~ f > ) rl f<<. <<If, vf >>r,  ! J 'I'. >>HI If H>> i<<f I f f( I>'l I, ff I ff << I'<<t <<.) f I<<r)w) 'ff>>$ <<>I>> ) r j II fir<< >>! II>> ~ I >>>> '> I 'f 'W>> I I AH I il r ~ t I' I I << H I 'll 4 If>> > 'ff<< Wr<< i I H>>>> y.p )<<, fl>>>> )7 I >>",I I 'I I>> H Notes: (1)-o One third of the yearly value given in Quantity of Solid Waste. Exposure Plant ~0 crating Personnel: It is estimated that exposure to plant operating personnel may increase by 0.5 man-rem due to the use of vendor solidification service. Safety and Technical Reviews: Documented safety and technical review in accordance with Technical Specification 6.5.1 have been completed for this modification. nnnnn 7/3 t' ,I c P" ~- l 'l ll ~ 'I N t N K II