ML19319B143

Forwards Environ & Radiation Safety Group Input for Rept to ACRS

ML19319B143
Person / Time
Site:	Oconee
Issue date:	07/09/1970
From:	Donna-Beth Howe US ATOMIC ENERGY COMMISSION (AEC)
To:	Long C US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 8001080931
Download: ML19319B143 (21)

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July 9, 1970 C. Long, Otief, PWR Branch No. 2, DRL IXIE POWER COW.'uY 0CONEE ACRS REPORT INPUT, DOCET NOS.

50-269, 270 and 237 Attached is our ACRS report input for Duke Power Company l.

0$onee Nuclear Power Plant, Docket Nos. 50-269, 270 and 287.

i P. W. Howe, Chief, Site r

Environmental and Radiation Safety Group Division of Reactor Licensing

Enclosure:

ACRS Report Input Duke Power Co.

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GCCCT ACRS EUPORT INP11T D00iZT.iOS. 50-269, 270 and 287 c.

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Site Location nnd Decerin fon-Ocotc+ 5tctiec !- is Ot.once County, South Carolina, about eight.

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-n.iles northeast of' Seneca,- South' Carolina. The site is adjacent

.to N.ake;Keouce which une forr.ed by impounding the Keovec and Little Riwrn with neparate' dans' and then joining the' lakes by l.

a canal, about half a mile north of the site. The nuclear station 1s about eight-tenths of a mile vest of Keowce River at the dam.

- Anderson, South Carolica, the nearest population center (1960

~populath.t. c.1 41,130), ir, 21.uilen a,uuih.

The applicant proposes sa minimum exclusion radius of one mile.- Based on the 10 CFR 100 definitionoftheexclusionradius,weconcludet$atthedistance aclectdd by; the applicent is acceptabic.

The ' applicant has proposed a.oi:: mile Lou Population Zone (LPZ)

- which he estimatec siill-contalit '3,400 people in 1970 and 8,900

- people by!2010. ~The trdnslent pcpulation within the LPZ is catimated by the applicant to bc 2,000 in.1970 and,-because of

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the development of. recreational and vacation facilities along Lake-

. Keovec,jis expected to increase to 19,000 by 2010.- Based'on the-I

. population distribution in the'. proposed LPZ and. the 10. CFR'100 -

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ofinition of the LPZ, we find that'the 6 mile LPZ!is acceptable.-

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i II. Geolorv and Seistnology.

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L; Plant structures will be founded'on ' Piedmont granite gneiss rock.

According to the'. applicant and'.to tlie AEC Division of Compliance,,

t no untpual problens concernin'g?'the foundation material occurred

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during const.ructiin. The geology and seismology of the Oconce site were reviewed in detail duri$$g the construction permit (CP) stage, and nothing has occurred togtlter our previous conclusions that the

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geological and seismologic'al conditions are acceptable for the safe

.- f operation,of the 'Oconce nuclear units.

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The Class I structures founded on bedrock were designed to withstand horizontal ground accelerations of 0.10g and 0.05g for the Design

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Basis Earthquake (DBE) and Operating Basis Earthquake (OBE) respec-tively. For Class -I structures on overburden a DBE acceleration of 0.15g was used as a design parameter.

III. Hydroloey Lake Kgowcc will be the source of condenser cooling water for the Oconee plant.

Cooling water will be withdrawn from the Little River arm of' the icke and discharged into Lake Keowee just vest of Keowee Dam on the north side of the plant property.

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In order to provide a continuous' supply:of emergency cooling water, a Class I submerged weir, 'which impounds 9-million cubic feet of water, has been constructeo across the lagoon frcm whicli con' denser

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. cooling water is withdrawn.

In the unlikely event that Keowce or Little River Dana should fail (both have been shown by the applicant to be capable of withstanding the DBE accelerations) the unter retained by the submerged weir would be circulated through the condensers and back to the intake lagoon providi ng continuous emergency cooling.

n.3 The applicant hes calculated that the Probable Ma :1 mum riood (PMF) 4 will result in a lake stage of 808 ft. :iSL.

Plant grade and critical plant components at 796 f t. }$L are provided flood protection to 815 ft. MSL by the Keowce Dam and intake canal dike.We have r$de an independent analyses of the anticipated wave effects and hav m...

e concluded that the seven feet of freeboard between the P:

stanc.and the 815 ft. !!SL protection level is adequate to prote t c

the nuclear plant against any credibic combination of wave effects and P:fF stage.

Based on the considerations discussed above, we and our hyd rological consultants, the U. S. Ccological Survey, conclude that the hydro logical conditions at Oconce Station are acceptchie relative t o the protection of public health and safety during operation of the Oconce nucicar units.

A copy of the U. S. Geological Survey report has been forwarded to the Committee.

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IV. !!cteoroloq

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During the CP review of the Oconce site, the staff and our meteorological consultants, ESSA, concluded that a " Valley" diffusion model wculd best characterize the meteorology of the site because of the complicated, rough topography between the Oconec nuclear unita and the nearest site boundary one mile to a~

the south. In this postulated model it was assumed that the c1vm effluent released as a consequence of a reactor accident would be channeled generally down the Kcowee River Valley to the nearest 4

site boundary. Co::plete mixing of the effluents was assumed to occur within the t.unflues of time topegraphic ridge: cieng r,ccecc 4

River between the reactors and the exclusion radius.

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-5 g c c,j,,3 Through marcorological diffucion factor was 7.4 x 10 his post CP meteorological program, the applicant,was to prove that the Valley model uns valid for the site.

The applicant conducted fif teen gas tracer (SF ) exp riments under 6

inversion. conditions, and in all cases the centerline concentration was lower than that which would have been predicted by the use of the equivalent Pasquill type diffusion conditions. The best agreement betvcen a measured concentration and. the concentration the Pasquill model would predict was at 680 meters where the measured concentration was a factor of 2.2 lower than what Pasquill categor-ization would have predicted, including building vake credit.

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e other' cases the meacured concentration was even a smaller fraction of the predicted concentration.

An examination of one year's data of the joint frequency tabulation

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of vind spccd, direction, and stability condition (delta T) data from a 150 foot tower indicated that nine percent of the time the diffusic"n rate was ecual to or wor c than Pasquill Type F and e2 1.5 m/sec wind speed.

This data indicates that a diffusion rate kh

.cquivalent to or vorac than Pacquill F and 1 m/sec wind speed occurc 5 percent of the time.

Ecc uce Of the height cf the trece in the vicinity of th+ natevi-ological teuer, the vind measurements were made at 150 fact-above grade rather than at a Icvel core appropriate for a ground level release (20 or 30 feet elevation). However, base'd upon a visual examination of the topography of the site, the 150 feet level appears to be a reasonable lower level at which to collect uind data and not have topographic interference.

In a related nat:cr, the applicant stated that a wind speed calibration check was r.ade in October 1969 which indicated the instrument vac measuring low by a factor of 1.4.

. llowever, there is no rigorous way to determine how long this situation 1had persisted and to what extend the data in the joint frequency tabulation were nffected. 'Therefore, the effcet of not correcting

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. c the 150 foot vind measure.~ents doun to the appropriate 20 or

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30' foot ~ clevation is prob-$ly compensated for by the 1.4 calibration factor.

For thin reason, uc believe that the unnodified 150 foot elevation uind data are n reasonable representation of the vind speeds.co be expectad at the 20 to 30 foot level.

In evaluating the radiological consequences of the design basis accidents, we have employed the usual staff model of Pasquill Type F and a vind speed of 1.0 meter per second (which was shown applicable by onsite data) with building wake credit and with an additional correction factor of 2.2 to account for the improved diffusion at Guvuue SLulion due to topographical eficces. Tnis results in a

-4 diffusion factor of 1.16 x 10 sec/m,3, while without the correction credit the staff's diffusion factor would bc 2.18 x 10 sec/m.

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ESSA, whose report has been foruarded to the Co Sittee, has concurred in this model.

V.

Environmental Radiation Surveillance A preoperational environmental monitoring program was initiated in January 1969, so that eso years of data would be available before startup of Unit 1.

(Water sampics from private vc11s and from the Keovec and Little River arms of Lake Keowce have been analyzed since 1966).

The preoperational program included the following sampics:

water,' airborne particulates, rain, settled dust, silt (river and 1-ama' t

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-7 lake), vegetation, aquatic vegetation, algae and plankton, fich, No environnental-radiation anomalics have been milk, and anirals.

indicated by the preoperational data thus far reported.

The operationni cnvironnental raonitoring program will be un extension of the preoperational nrogram with the follouing additions in order to provide a more comprehensive progran to quantitate the environ-mental effects of operating the nuclear units:

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cro additional onsite air monitoring stations, a continuous water sampling station on the Eccuee River just 2.

uithin the exclusion radius, and a thermoluninescent dosincter network within the one mile 3.

exclusion radius.

The frequency of sampling and types of analyses of each nedia are given in Table 2-la and in Sections 2.7.2 and 2.7.3 of the FSAR.

These references vill be incorporated in the Oconce tech specs.

Duho Power Coapany is cooperating Lich the South Carolina State Board of Health, South Carolina Pollution Control Authority, South Carolina Wildlife Resources Departncnt, and the U. S. Fish and Wildlife Service in matters concerning the environment.

The Oconee environmental nonitoring program enconpasses the recommendations of the U. S. Fish and Wildlife Service with the

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exception of campling within 500 feet of the liquid effluent outfall.

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1 Die applicant has added a commitment to nampic aquatic biota, 1

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crustaceans or molluscu, benthic organisms, cnd botton sediments as near the outfall as they. can be found.

The applicant has stated that the scouring effect of the Knowee llydro Plant discharge will prevent the developnent of there organisms or bottom sediment close to the outfall. The comments of the Fish and Wildlife Service have been fodiearded to the Committee, w.

"A 1 Based on the description of the environcental monitoring program in the PSAR and FSAR, we conclude that the environmental surveillance for Oconee Station is acceptable.

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Radioactive Waste Mananenent As stated in the FSAR, liquid radioactive vastes arp segregated in rocciving tanks according to their source, sample,d, cnalyzed and then treated.

Based upon the analysce, the liquid waste will be treated 'by one of the following methods:

1. - discharge to the Keouce llydro Plant tailrace; 2.

holdup for decay, resampled and analyzed, and discharge to the Keouce tailrace;

3., concentration by evaporation with ultimate disposal as a solid waste.

Also, according to the applicant, liquid vaste will be diluted, 09 necessary, by t!ie hydro plant discharge (30 cubic feet per second to JIT.1 g

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119,800- cfs); to'.meetithe concentration limics of.10 CFR 20.

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in order; to retain operational flexibility,- the applicant only assumes the ininimum dilution (30 cfo) in estimating' the annual-release. limits.

i/fcording to the applicant, the Keowce liydro Plant, which is controlled

.from the 'Oconee nucicar plant control rvoms, is expected to be operated

at least ucchly,lif not on a daily basis (tech spec scetion 15.4.4).

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n'11Ius,,[f(ow cubstantially greater than the minimum leakage flow uill L usually be available.

O In the applicant's design evaluation of the Oconce liquid radwaste -

treattent system. a ninimun holdup time of 30 days was assumed.

This is indicative of the1 amount of waste stornce ennkaga thnt is available in.the plant: j8 100 ft for.the miscellaneous liquid waste and

66,000 ft ~ for the primary coolant (coolant storat;c. system).

In the

- proposed tech spec (nection 15.3.9), ' the applicant has commtitted to

treat all 11guld radioactive waste "... to reduce the quantitics released to as low a icvel as practicabic if the concentration' upon dilution with nornal llydro Plant Leakage at the point of release would g.

be greater :than 1/10 MPC vithout processing." Tac. staff position' is -

thL:.:allL radioactive-liquid waste should be trected with the equipment-7 provided--in; this -. case c the evaporator. ;In additionf uith the available holdup l capacity [in:theOconceplant2and' with the -e <pected operating routine' of the Kcowee liydro Plant,- we conclude that-reasonable effort d

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.t on the part 'of. the-' applicant dictaten that the liquid. radvaste be 2c. d

.I-discharged'only if.the hydro plant is being operated.

Thus, the waste ~ would roccive a much greater instantaneous dilution than if

< the hydro plant uns not operating.

' The icntire - radwaste' sys tem is. located ' bclou grade in Class I structures so' that dn the event of an accidental spill,- the liquid. vaste sill be N!

. retained within the structures.

In order for accidental discharges bs*!

to the environment to occur, vaste vould have to be pumped from the S

below grade storage tanks to the Keouce tailrace through a discharge

~ valve uhich'is closed by' a high radiation signal from a radiation monitor.

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.The reactor coolant treatment system, which is provided to recover a

boron' and to purify the coolant, provides additienal _" waste treatment"

- of the primary coolcnt before it reaches the liquid waste disposal-system.: The, coolant treatment system, which operates on a batch basis, receives' liquid: from the primary coolant bleed holdup tanks through~ bleed evaporator denineralizers L(deborating domineralizers).

' The coolant. is ~ fed to the - coolant bleed ev'.porator, 'and. then the condensate:is sent tio the condensate test tanks.. At.the operator.

. option the testi-tanks h contents can be_ (1) routed to ' the vastie feed.

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, tank. (2) returned to'the coolant biced evaporator feed tank, (3) _

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g returned to the-coolantbleed holdup _ tanks, or (4) released to the O

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liquid waste effluent header.

In conjunction with options 3 and 4, the tect ' tank ef fluent can be passed through the condensate f

detaincralizern.

The solid waste disponal cystem includen eqttipment to collect and store two years' generation of spent denincralizer recina and a hydrauli: prenn for use in handling compressible solid wacte. All solid w ste are ultinately drummed and shipped offsite for final g

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-disposal.

l The gaseous waste disposal systene are conr. acted to vent headers which collect potentially radioactive of f-gases fron all components unich may contain radiogases.

Ecfore rc1cese to the environncnt, these off-gaces are processed either through a waste gas exhauster and a filter -bank composed of a profilter, an absolute filter and a charcoal filter, or through the unste gan decay tanhn and then through the filter train. The gan decay tank contente are satpled and analy cd prior to release to the plant vent. Units 1 and 2 sharc n gar.cous waste disposal system and Unit 3 has an independent system.

houcver, these systema can be interconnected through double isolation valves between the respective vent headern; thus, operational flexibility is provided in the event one waste gas systen is temporarily out of service.

According to the TSAR, the waste gas exhauster will be used when large

, volumes of gas containing littic or no radioactivity are to be released.

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The exhaustera(fan) and the isolation-' valves on the waste gas.

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exhauster 'and decay tank discharge lines arc interlocked with a

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1 radiation nonitor no. that in the event of-high Icvel activity, n.-

idischarge"throunh ' these paths will bc - ter:ainated.

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decor [ ding Lto the ' applicant, containcent purging vill be through m

particui" ate (11 EPA) and-charcoal filters to the plant vent. However,.

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iunder conditions of. low level activity in the containment air, the applicant proposes'to bypass the. filter' system.

iIn' the proposed tech -cpecs :(riection 15.3.9) the applicant states that ".... ;;cccous cactes.uill be processed prier te ' rcicasc to rcducc r,.

the quantity relccced to as: low a 1cvel as practicable if the conecatration at-the point of release sculd be.gre'a'ter than'1/10 HPC

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. with'out: processing."

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.In the technical specifications we shall require that gaseous radio-

active waste passing through the off-gas system be hcid. for decay

.(the holdup time will be establisticd in the tech specs) and that all s

radioactive gaseous vaste be filtered through the systems previously

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f idesignated a restricted larca which is basically the confines of the 1

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,t enuclear pla'ntistructures..He also desires to assume an-atmospheric w;

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dilution factor which would be applicable to the one mile site i

exclusion radius in establishing the Oconce gaseous release linits.

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The applicant believes that 10 CFR 20.106 provides the basis for this approach. IM clairs that he can prevent occace originated radiation exposure of the public vithin the exclusion radius by venting from the decay tanks only during periods of favorable atmosph ric diffusion conditions. The applicant has taken this position because:

(1) there is a visitor's center approximately 300 r:cters frun Unit 1, (2) construction will be continuing on Units 2 and 3 af ter Unit 1 begins operation, (3) there are tc:7.:rar, ccnstructicn vorhers' campn within the escluaica radius, and (4) a portion of Lake Keouce within the exclusion radius will be used for recreational purposes.

The staff position is that the annual release li[its should be calculated at the nearest boundary of the designated restricted arca uith an atmospheric diffusion factor appropriate for this dictance..

-Based on the descriptions of the raduaste syster.2 and the data

,provided in the PSAR nnd the FSAR, we conclude that the Oconce rad-waste systems are capabic of providing waste effluents which can be considered "as low as practicabic."

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F Accident'!nalyses:

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.The ' staff hasianalyr.ed the radiological consequences of the

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i folloding designj basini occidents. in evaluating site acceptability:

-thcl loss'.of coolant accident, refueling accident,'g':: decay tunh a

' rupture, ' nteam 'line. break,.stean generator tube rupture and rod

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ejection' accident. The dose conocquences are given in Table 1,

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- c an ? the' various. accident 'assunptions are" provided in Tcbic 2.

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dose conscquences for' the steamline break, stena generator tube

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rupture. and. the rod cjection' accidents represent. only the' prinary systen contribution. The additional contribution du'e to relesne of' moe.nn.iary sista f oet tvitv will be reported later ef ter the _e_d hoc ep.

co:mnittec on accidents ' involving the secondary system has completed its uork.in. developing standard staff models.

This*is expected before the middic of' July.

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TABLE I

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• MDIOLOCdAL CONSEQUENCES OF DESIGN BASIS ACCIDENTS

'PJO Il0Un SITE EOUNDARY DOSES COURSE OF ACCIDE!;T LPZ DOSES

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AT 6 !!ILES:RE!!

ACCI DE'iT TIIYROIn pl? OLE BODY

, THYROID W1? OLE BOD'l LOSS OF COOLANT -

190 2

200 I

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REFUELING (with filters)y 30

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<l Mr GAS DECAY TA"K RUPTURE 2

<1 STEAM GENEMTOR2 TUBE RUPTURE 65 1

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<l STEldI LIII. LitEAE' 1

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ROD FJECTION~

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We have informed the applicant that charcoal filtration c<f the spent fuel pool area atransphere is required durin;; fuel hr.r.dliur,.

lie has not as, reed to this.

2 Primary system contribution only.

Secondary cystem co nponent will be evaluated following g hoc cor= ittee's developmant of a staff model.

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. LOSS '0F COOLMiT ACC3Et.T - (IDCA) ~

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.A..' Reactor power icvel is 2568 We.

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( B.' 100% of[the core. noble gaslinventory is released to the contain:nent and is Taveflable for ' leakage to -the environment.

C.-

50% of the core lodine inventory is relcaned to-the containnent.

-D.7 50%~of the. iodine plate.cout in the containment.

E.' ~10%.of rensining iodinc is in the organic form.

Conticinment 'dcaign leak rate is 0.25%/ day.

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After the first day, centainment 1cakage is reduced to.45% of.

design valu'e due to pressure reduction.

-H. - 50% of. containment leakage is through the penetration room and'

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ite Hhi% and en.,ren n t ffiters.

I. :50% of containc.ent leakage is released to the atmosphere unfiltered.

J.: Charcoal filter efficiency for iodine is 90%.' '

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Breathing rate is 8 hrn :

3.47x10 m /sec *

-4 3 24 hrs : 1.75x10 m /sce.

-4 3 1-30' days: 12.32x10 m /sec.

L.-7 Meteorology:

- Pasqtiil1~ condition "F" Windspeed a.1.0 n/sec tierrain correction factor = 2.2 building wake effect:

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' ground level release.

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Exclusion radius.is 1 mileb m

N 1 Low 1 Population. Zone _is 6 miles.

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REFl!ELIMG ACCIDI:::T A.

Renctor power Icvel in 2568 Mut.

B.

208 pins (1 fuel bundle) are damaged.

C.

20% of noble gases is releaced from the damaged fuel pins to the environment.

D.

10% of the f odines is released to the fuel pool water.

E.

Decontar.1 nation f actor for iodines in the water is 10.

F.

Charcoal filter iodine removal efficiency is 90%.

G.

Radiil peaking factor is 1.68.

II. Accident occurs after 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> fuel decay time.

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Breathing rate is 3.47x10 m /sec.

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Duration of accident is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

K.

Metco ology - see LOCA.

IIl. CAS DECAY TANK RUPTURE A.

Entire nobic gas centent of one prinary coolant volume (11,830 f t.

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is in the decay tank prior to rupture.

B.

Source term is given by applicant.

C.

Entire contents of decay tanks are released to the atmosphere in two hours.

D.

Average decay energy of fission products released is 0.7 Mev per disintegration.

E.

Meteorology - same as LOCA.

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MOD ET::CTIO'i ACCIDi" 7 A.

Reactor power Icvel is 2568 Mat.

B.

4.1% of fuel undergoes cladding damage.

C.

Prior reactor coolant fiscion product inventory is that given by applicant as his design cource term.

D.

20% of noble gasec r21 cased from damaged fuel.

E.

10% of iodines released from danaged fuel.

F.

Release path is from primary cyctem to steam generator then to environment through air ejector - assuming a 10 gpn stcan generator ge,-

tube Icah.

G.

Prieary coolant volume is 11,830 ft 11.

Dur., tion of accident is 1.7 hrc.

I.

Breathing rate is 3.47 x 10'*n /sec.

J.

Meteorology: See LOCA.

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Steam Generator Tetc 11unture A.

Reactor operating with - ficcion product inventory given by applicant as his design source term.

B.

Exiaiting 10 gpn steam generator tube Icak.

C.

Accident duration - 30 minutes.

D.

Voler.e of primary coolant' lor.i.

1,980 Ft E.

Prie.ary systen volume is 11,830 ft F.

All l'odines and noble gases released to steam nensrator are exhausted to the environuent uithout a partition f actor.

-4 3 C.

Breathing rate is 3.47x10 m /sec.

11.

Meteorology - see LOCA.

I.

Sec Part Vil for assumptions used in deriving secondary side

.s activity relcace.

VI.

Stenn T.ine Bresh A.

Duration of accident is 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

B.

Steam generater tube leak is 10 gpm.

C.

Primary coolant fission product inventory is as given by applicant as his design source term.

D.

Tube leak is constant for term of accident.

E.. All iodines and nobic gases that carry over to secondary side are released to the environe.cnt.

-4 3 F.

Breathing rate is 3.47x10 m /sec.

G.

Metcorology: See LOCA l.

11. See'part VII for assumptions used in deriving cecondary

~

side contribution.

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e c,

VII.

Secondary Side contribution to Acet<lents_ (IV-V7 To be cubnitted later folicving conpletion of choc conr.ittee's vork on standardizing the assu.vtions for the, c.an generator tube the steau line break acciden: cnd the rod ejection rupture accident, accident.

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