ML20071F411

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Forwards Draft Responses to Radiological Assessment Branch Request for Addl Info & Draft FSAR Changes Resulting from 830428 Telcon W/Nrc.Schedule for Formally Incorporating Changes Into Future FSAR Revs Noted
ML20071F411
Person / Time
Site: Limerick  
Issue date: 05/16/1983
From: Bradley E
PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To: Schwencer A
Office of Nuclear Reactor Regulation
References
NUDOCS 8305180432
Download: ML20071F411 (55)
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Text

{{#Wiki_filter:! .e 1 i PHILADELPHIA ELECTRIC COMPANY 2301 M ARKET STREET P.O. BOX 8699 PHILADELPHI A. PA.19101 l CDW ARD G. E AUER, J R. N.Eu (FUGENE J. BR ADLEY assoceATE eEssanAL couNsE6 DON ALD BLANKEN EUDOLPH A. CHILLEMI g E. C. MI R K H A LL T. H. M AH ER CORN ELL PAUL AUERBACH ASSISTANT GENER AL COUNSEL CDW ARD J. CULLEN. JR. THOM AS H. MILLER, J R. IRENE A. McMEMM A assistant COUNSEk bir. A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing U. S. Nuclear Regulatory Commission h'ashington, DC 20555

Subject:

Limerick Generating Station, Units 1 and 2 Request for Information from the Radiological Assessment Branch

References:

(1) A. Schwencher to E. G. Bauer, Jr. letter dated March 11, 1983. I (2) Conference call between the Radiological Assessment Branch and Philadelphia Electric f Company on April 28, 1983. File: Gov't 1-1 (NRC)

Dear Mr. Schwencer:

j The attached documents are draft question response changes and draft text changes to the FSAR resulting from the discussions with Mr. Mike La Mastra, Radiological Assessment Branch reviewer, at the referenced conference call. The changes to items 3 and 9 will be formally incorporated into the FSAR revision scheduled for May, 1983. The balance of the changes will be formally incorporated into the FSAR revision scheduled for June, 1983. i Sincerely, 8305180432 h h 2 h PDR ADOCK o PDR A wA i ' ' 'f( E.. r LN/cw/G-1 cc: See Attached Service List

?f * ..h4 ge Lawrence Brenner (w/o enclosure) d cc: . Judge Richard F. Cole (w/o enclosure) Judge Peter A. Morris (w/o enclosure) Troy B. Conner, Jr., Esq. (w/o enclosure) Ann P. Hodgdon (w/o enclosure) Mr. Frank R.~ Romano (w/o enclosure) Mr. Robert L. Anthony (w/o enclosure) Mr. Marvin I. Lewis (w/o enclosure) Judith A. Dorsey, Esq. (w/o enclosure) Charles W. Elliott, Esq. (w/o enclosure) Mr. Alan J. Nogee (w/o enclosure) Thomas Y. Au, Esq. (w/o enclosure) Mr. Thomas Gerusky (w/o enclosure) Director, Pennsylvania Emergency Management Agency (w/o enclosure) Mr. Steven P. Hershey (w/o enclosure) James M. Neill, Esq. (w/o enclosure) Donald S. Bronstein, Esq. (w/o enclosure) Mr. Joseph 11. White, III (w/o enclosure) -Walter W. Cohen, Esq. (w/o enclosure) Robert J. Sugarman, Esq. (w/o enclosure) Rodney D. Johnson (w/o enclosure) Atomic Safety and Licensing Appeal Board (w/o enclosure) Atomic Safety and Licensing Board Panel (w/o enclosure) Docket and Service Sectio'n (w/o enclosure) i 4

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OP d M~ud 23 LGS FSAR ((^% d 2. High-radiation alarm light (amber) 3. Downscale alarm light (white) 4. Alarm reset (push button) 5. Meter zero adjust (on the amplifier) 6. Alarm level adjust 7. Trip check pushbutton 8. Power supply switch and " power-on" light (clear lens) 9. Indicators to show power supply voltages 10. Annunciator outputs 11. Recorder outputs e. The radiation monitors are calibrated at regular time intervals in accordance with station procedures. Calibration methods are covered in detail in the equipment procedures manual. f. The following annunciators are located in the control room to alert the operator: 1. Reactor enclosure area, high-radiation (Units 1 and 2) 2. Refueling floor area, high-radiation (Units 1 and 2) 3. Turbine enclosure area, high-radiation (Units 1 and 2) 4. Turbine enclosure common area, high-radiation 5. Radwaste enclosure area, high-radiation (common) 6. Refueling hoistway common area, high-radiation 7. Hot maintenance shop, high-radiation (common) 8. Unitized area, low-radiation (trouble) 9. Common area, low-radiation (trouble) l 12.3.4.1.3 Local Area Monitors i i In addition to the area radiation monitors described above, a I total of four local area monitors is provided, located on each of the two refueling bridges and the two turbine enclosure crane cabs. The essential differences between these monitors and those area monitors described above are as follows: l l a. No outputs to the control room are provided. b. Alarms are local only. c. No recorders are provided. d. Local power (battery) packs are provided in the event of external power cutoff. l W thy. anerw g nrtu v.& r w:// be ayA in -flu dewil 6: piv.+ p r a n<l . m co, ,{ &, y.ca,y stoorlding-

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..' ~ ) ' LCS FSAR INTRODUCTION & CENERAL DESCRIPTION CHAPTER 1 FIGURES Title Figure No. Piping and Mechanical Reactor Bldg. Drywell Unit 1, Plan at Elev. 310 Feet, Areas 11,12 1.2-70 15 & 16 Piping and Mechanical Reactor Bldg. Unit No. I 1.2-71 bection B-B, Areas 11, 12 & 16 Piping and Equipment Layout, Drywell 1.2-72 Unit 1, Section D-D Reactor System Heat Balance 1.2-73 Piping and Instrument Diagram Legend 1.20-1 1.10-2 Logic symbols r-y .l3*l A<LCSS ?"%S f I _r- , ; _ _ _w I l'af e K e oca e s n Focier.s ist ECLS/AtHL

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The shielding des review will include the fo wing areas a. Main Contro oom 0p[h b. Technical Suppor enter 00 Post-Accident Samplin rea p c. l ~ d. Post-Accident Sam e Analy ' Area $N } fgIN e. Secondary Co inment t k f. Securit enter g. Rad te Control Area h. 2, emote Shutdown Panel Area l l zoactive source terms used will be equivalent to the1.7, and The re terms recommended in Regulatory Guides 1.3, 1.4, Existing shielding for the Main Control Room is sou 15.6.5. m 1.13-24 Rev. 3, 03/82

n. ] gy lW/7h Y~ 7 eetions 6.4.2 and 1 nts is described ( Oualification A A + e II.B.3 pOSTACCIDENT SAMPLING l Position A design and operational review of the reactor coolant and containment atmosphere sampling line systems shall be performed to determine the capability of personnel to promptly ob a radiation exposure to any individual in excess of 3 and1 Accident conditions should assume a Regulatory Guide Consequences of a Loss-of-Coolant Accident for Boiling Radiological Consequences of a Loss-of-Coolant Accident for If the Pressurized Water Reactor" release of fission products. review indicates that personnel could not promptly and safely obtain samples, additional design features or shielding should be provided to meet the criteria. A design and operational review of the radiological spectrum analysis facilities shall be performed to determine the certain capability to promptly quantify (in less than 2 hours) radionuclides that are indicators of the degree of core damage. Such radionuclides are noble gases which indicate claddingThe initial ( failure and isotopes which indicate fuel melting. reactor coolant spectrum should correspond to a EsgulatoryThe review Guide 1.3 or 1.4 release. effects of direct radiation from piping and compon auxiliary building and possible contamination and directIf the review indicates radiation from airborne effluents.the analyses required cannot be per existing equipment, then design modifications or equipment procurement shall be undertaken to meet the criteria. In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions. Procedures shall be provided to perform boron and ch (Regulatory Guide 1.3 or 1.4 source term). be capable of being completed promptly (i.e., the b shift). ( Rev. 3, 03/82 1.13-25 --e.-.-.%,,,__ w ,m-- ---w--- ---m--,-,n,.

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/e k.. Sn+roductith 4# m desyn review of plant diel R' R A 7 In 8[/ S i[CD. h A i n n s ,yp4pr ftl(jfE{r 0 7 O,.TfC m f. f. ) 6 11$ l.S ~ Mr Me $f YbC-f0 V' BW NO I lAlf pywl uIfn1611f arens A wh m zy reg u!r e-p,,4kapreuem the development of special post-accident procedures, installation of additional permanent or temporary shielding, relocation of components or piping, or requalification of components. - T Areas and equipment which are vital for post-accident occupancy or operation were evaluated to determine if access and performance of required operator activities or equipment functions might be unduly impaired due to the presence of the postulated ra.diation source in the selected systems. Systems required or postulated to process highly radioactive fluids or gases outside the containment during post-accident conditions were selected for evaluation. Radiation levels in adjacent plant areas due to contained sources in piping and equipment of these i systems were estimated. b m d @ rddf CA# M h pa"1 lcaQefrom e C&nfainment omd f" exre aho d1Clun'ed aeident.souttet Y as E. saatuano8, %:,.a:m deza bed :. = aSam g;prfain'j c aeowicason 6 dwarau e ne a+td a.sammo[y ef % inethcdologsuch'fc a'eferm fed}sfion dosel & Yhe.re areas amd Yo c4ayment' 1rverr a fitcA1 p?ge> drefreseAted be(04),W11e iden'fifr'cafion W essenh'ai uynent amd ' tite ecuts d q redew ef yupment-kr-A postulated the --.-w ,-,y ee m-. w

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e) [ Vital dres fdesti6caffm Areas which may require occupancy to permit an operator to aid in the mitigation of or recovery from an accident are designated as vital areas. A review of the Limerick Generating Station was made which determined that the following areas should be designated vital areas. Continuous Occupancy k t* l 1) Main Control Room 2) Technical Support Center l 3) Operations Support Center l 4) Security Center J_nf requent Occupane_y l 1) Counting Room 2) Radiochemistry laboratory f t scalem5

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,--4.-,g ,7 4) 5) Hi'AC panels at el. 304' 6) Radcaste control room ~7) D iC.S c l g C m W a fspy 4t y f x Potenfal VHa/ areas that-are not- /)sted above edere excfuded isv 1he .f9// swim reasons T4e post--Lecn (reambmeh

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c+nfro/g gstem,,confainmemt isolafon s hydrogen Aca%id enreg5m.p+ow 9 5 (44esob ~ ~ ~ ~ axe an c er fsis/9(SW -.pu hs remofeyl C e m fro / W G 'fA C tpferakt in Yhe mak confrof rnm 42md repse no Iocs( acess. The<e is no mamaa/ &cs Glip ment strcA. At Omedek. fnStrament (2w'IS v d

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,rwtp< anrtro/ cexters nof mCladaf becau.ce 'the o,sd,tofC& ate f 'ms confrol aMd eIfanieat d eued l ,p & a ( ene es4 %asnD puw.dn #emath conYrol r6+m amoc IC Acc.opf ehed Ao air & re irk eeny /ccol adr%, G 77sred-tx@ t Li ittS W S fzrf 439'd 71 'fh a f r M/e w e also ixenH6 & metadat'. op 'rsi* review. Yf'EA1WAflor/WK)hI W/f SOC'4A&40 f d C. 5Clt&Vn Cf W6ms ) hl Qgg A re.iew M t': if-'"C:....;

" -^- was made to determine which systems could be required mfd 48 to operate and/or*Txpected to contain highly radioactive materials following a postulated accident where substantial core damage has occurrei.

The results of this review,are presented below.Afecter tl;rrt 7.sohntoeyr (,6Clo.otf @ h 9t

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Core Spray HPCI),7RCIC), M /tsidus/ // Art k %~ C m ps!st(Kht), pn/.fdfey uead $})dsge Fill &ternt S. 9 4The Core Spray, RHR, ITPCI (water side) em! .T \\A RCIC (water side ysystems would contain .h suppression pool water being injected to the reactor coolant system. Although the HFCI and ROIC systems could 'so carry condensate, 4 (~ assumed for this suppression pool water The steam sides of review for conservatism. the HPCI and RCIC systems would operate on t l reactor steam. ,2o 3 ST'W1 U @y RHR (Shutdown Cooling Mode) The RHR system recirculates reactor water when it operates in the shutdown cooling mode. Before operation in this modo can be initiatei the reactor must be depressurized to less than 75 psig. This depressurization is expected to remove substantially all of the noble gases $t96. W released into the reactor water. Following Ipppy a postulated serious accident the HPCI, RCIC, M &f JI and Core Spray Systems would g(gf ed,' '#@l RHR (LPCI mode ), inject water into the reactor c This water from the condensate tank and/or the g/d.5SIgf g() suppression pool would dilute the reactor waterprior to the in g the RHR system. This edrieMung review emute g gtatt. jassume.ufthat there are no noble gases in the f reactor water in the RHR system for the shutdown g "r-r rr. rin:4 t'. acc r t g ooling mode _ sm. aih im s ;; .? 2. h t.... J ;;..

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) determined that yP postulated accident. Prior to a scram the CRD housings It they will not. contain condensate water delivered by the CRD When a scram occurs some of this conden- / pumps. sate water from the CRD is discharged to the scram discharge header. After the scram, some condensate and reactor water flows to the scra-l discharge header until it is completely filled. This takes a matter of seconds. Since the vents and drains in the scram discharge header are isola-ted by the scram, all discharge flow then steps. Since it is not reasonable to assume that signifi-cant core damage occurs in the first few seconds the scram discharge header ~ following a scram, will contain only a mixture of condensate and pre-accident reactor water following this post.nlated accident. teon iMu asamq(twcd) 4, % amot System For a major accident with resulting core damage, the RhTU system would be isolated and would contain no highly radioactive materials beyond the second isolation valve. On a Bh*R this system needed for reactor coolant system venting. is not It would not be practical to use it for accident recovery af ter a major accident. % therefore assuned that this system would no operate with highly contaminated reactor water. } fg Gaseous Radwaste System For a major accident with resulting core da age, it would not be practical to use the gaseous Noble radwaste system for accident recovery. gas isotopes with long lives would cause exces-sive offsite doses if the gaseous radwaste after a design basis accident. system g n a w v,h m m s, a nn,+< ~ casaid rt M 9)or-ra t e. Auiaent Samyhj [urer Mr ( 391 jda used affBt w L26dd W

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44l$t{d !! frinter ( CSW4tinmerttgas, ' confainnsenf gas, Jen:Brtfor reacfw Greferst restosted pr Qarcheed) ),, eru er s w casion,l/n5,w w. a wl W A & 6 V1' PrL/. %e Jdenf9/My j 7.s Mtokumoer sttnrospheric Mfrot Jyte~s I The recoSbiner syste:ravould recirculate pri-arn M, containment gas aftef a serious accident in order

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to keep hydrogen and oxygen concentrations at 4 -62 acceptable levels, w/ ge</,4c Y3 NN#I'W EF'CIO3"!* Jtandb Geu TttAnot Tem K sk. a.y .: z= <<ta prs y m fgEls} [ .sGrs a.wt /EJt eyrt - 2n' "---tM The r:rdt., 0:: % ='*=-+ a In:::: cre ' -'--it M. :, :;1 would colle ct airborne activity in the secondary containment following a serious accident. Radioactivity would be collected on the f11ters and charcoal beds in these systems. l 9. Containment The free. volume d % psnog containment mifia} certfaist large amounk is azumei fo &pc -aakde,11 - aesy. Thee Jmeces,as wen as nee asiarn ed fer-the. styptraften fsoI, are deschbed belous Ae +Argh 111e dywdi amd wetryef uld//s wGu(d cause a. gty e increae k na sea,1s'a<y W Con /ammemt airherne ansdpipy 4t'Wes omd #erefore tau not thcluded th #h=< ' p revie ). ,-wv-v~"v~ e e,~ ~~

$..$0ute e f6ltt216 hAek9913 ( The following release fractions were used as a basis for determining the concentrations for theyshielding kgapsf.o.a Awp review Source A: Containment atmosphere: 1004 noble gases, 25% halogens .,_..J p 9u h -. s & __ r. 2:;;r:_.!;;." 1 _ -_ Suppression pool liquid: 50% halogens, .y Source /g: 14 solids Source p/-- Reactor [ team: 100t noble gases, 4 f: 254 halogens These release fractions were applied to the total curies available for the particular chemical for

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I an equilibrium fission product inventory for a Ressuiaed <ac% actaat : !00% "W' "$ light water reactor core. s-o% halo ens, .swee o: 9 %.50 AS E. Seuice Teem Mate /s factions rjie assumptions caet for release nounroes A~* ~ id{lf/ \\9 $ 11 116V d4.0 M C Cg O(J$1M6$ $YC b hese release fractions are, however, only the first step in modeling the source tern.s for the activity concentrations in the sytems under review. The decay time and dilution volume also af fect the rationale for the selection of values for these parameters. Decay Time eas taker. For conservatism, no decay tirre credit I for the radioactive decay that might o:cer befcre fission products would be transported to the various systems.

r v 4 Dilution volum2 ( S. The volume used for dilution is inportant, affecting the calculations of dose rate in a The following dilution vola'es linear fashion. were used with the release fractions and decay the final searce times listed above to arrive at terms for the shielding reviews. Drywell and suppression pool free dE ' Source A: volumes.

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[ bud en tea,& C 4.,,,,.L _ .,4. (20l06 AMI L.: ;.12: :.:. e m era} a The volume of the reactor coolant Sourcef: plus the suppression poo1AldfW .f pg g. frdMW d.) (, The total reactor system steam volume. Source pf: 1)nf. Volume d Y$gg. pgat}gy geofen atssicos. G .$&urse. Qt e,.: .,,, e n,. c..t ,3.} . aW.toonai '.S e c w c e.sOrgwet( & skk.ecc.$ T;~'- "Ti.. ~...-. t&Josn n c &.Secondnv (&ntcanenc.y& fuudent operehnj recdc.t were a.numaf fer &ck Sfem. limits of the connected piping In defining the subject to contanination listed below, norma 31y ' n i<1 sho*. valves were assumed to remain shut. p Core spray system - Source / 8 \\ g,dj lI J t High pressure coolant injection system f Liquid - Source g8 Source PC(with credit for stea-Steam - spe.ific activity reduction due c to turbine operation). . Reactor core isolation cooling system Liquid - Source f3 Source g(with credit for stea? Steam - specific activity reduction due to turbine operation). Pesidual heat removal system . Souse. 8 (p// MiCdu) _e,.,,., n ,. e y..u y.. .s ..w. .-4_ -__;_ 7 h

6 /?cadenf-Jamptig lines QQQ fest seuice. M i &a.s sompte noes ti usk. compte una Jearce c contmhmerrt Mimosphede fortrol(/'ecembider) 3 (eqwer free 3.stesn sou ec e A volume &l) y D w et/ Swice A ....Sta+1/G gas 77eafmerr/ Jytem, A*eacYe>- 51clo.ru?e f e c s e c a la ti M Jarte+,,s n>w( q 38%& &N/ rVt09tY /YC$V$ I The following major assumptions were used to ~'l f ~ calculate theamirborne radiation doses and ~ j the radiation ~doseh for the SGTS filters and the RERS filters. \\ er.;8MdMy CAMfC<inm M1~ - .s d, Tp 100% of the noble gases and 25% of the ~~ halogens are available for leakage into the secondary containment. b 4 The primary to secondary containment leak rate is 0.54 per day. C,I,, Airborne activity in the secondary con-tainment is confined to spaces below ca.gan,wsa#F the refueling floor. w f g, g The RERS flow rate is two ir changes i per hour. ci,,g.L4a ? -~ air g, k The SGTS flow rate is one halfj change per day. The RERS charcoal filter is 954 efficient [, ff with respect to halogens. I g. s .**6

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} 72e.s6rs c41 moat A//cr ).s woroe A)cet ,41 . cay 1r rapect Yo /t*/oj*ns. & A/hsr... .__ fnx < h J n. 9 ~ twt 1 M~-. c -- - < n. cr - < - - w ~Anns, >>$ W. 'Je _; ' L, s _s__ _ -c' r ciCi & +t<.f d s. i dMe,_ yes ( rtCa rr Wrg h, The activity inventory in the core was based on 1000 days burnup and daughter product formation was not considered. These assumptions, which have off-setting - ~ - - effect.s, were necessitated by limitations - - ~ ~ ~ in the computer code used to treat the secondary containment. l transport of activity from primary to L,s. w gasnes et w eets amd urs .C Ifets... One .sufh5 en V 10 AUSYO th......- . CAea+14y & %e duraf7>+t 9f Y11t. -.act)d&nf. 1'. ifi<berit e. %sa Vibr/ Hreas M/ vhe ikI areas arth acsas 4 d,.... tzed C are. focafed sr Y1se b<rhu e ericlosute, radu20 tfe ertcfOsare, c6n fr0(._..... -. .skucdwe, adhiinisfraf7bn hutl6 tin . tah eicai .s u.pyor7-costfcr. 111e Yramspert l ~ ~ h

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siit e grated gamma dose was then determined by adding. _

the post-LOCA integrated dose to thegnorma2 w 40 f4 operating integrated dose. The equipment'quali-fication levels for all safety-related electrical components were compared to the applicable calcu-lated dose. For those components initially listed as inadequately qualified, more detailed ~ calculations were performed, taking into account dose / distance relationships in order to determine more realistic doses. A set of dose / distance curves for each system was developed as part of this effort. ~ ' ~~ j I -G

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Q&N 5 'L f 0 f Both gamma and beta 4 oses were calculated I d for the primary containment. The doses 6 were calculated by assuming that 100% of k the core noble gas inventory, 50% of the core halogen inventory, and it of the core solid fission product inventory are re-leased. These source terms aret consistent with those specified in NUREG-0588 and NUREG-0737. The primary containment airborne dose calculations assumed that 50% of the 50% (i.e. 25%) halogen release from the core plates out instantaneously, as assumed implicitly in Regulatory Guide 1.3, Rev. 2. The airborne doses were calculated assuming source terms diluted by the primary contain-ment (drywell and wetwell) free volume. These assumptions are consistent with those specified in NUREG-0737. The beta doses and dose rates were calcu-lated assuming an infinite cloud geometry. For components inside primary containment, the total integrated gamma doses were calculated by g$yttd adding the post-LOCA primary containmentVifloud g dose to the 40 year normal operating dose. tese distance relationships were not used to reduce # post-LOCA doses inside primary containment. A hfJ /N1 N6f AW $d27S &r1YO 1/ Me} - A'e' lafat' E artest br speb'Aed 6 Weal Ga8& 'the 4hses .sec&rta'a# Shr1-abraent aiere n(so as n ked er t~ hAarfion Cal 6uIa eQ'A e e ftwpec-K o r Y1ce. 7'S ent um-ptm ent, Wte-p- L0 ft annia a'oset i.s Wct confact sf.'ae d e.5675 '8061s. )Qw-Y)id r C+7tb m Arf S Y e d s., % e

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n. P h f d f.5 9 h ts d l CSkfC bhntll$ s The general basis for personnel radiation exposure guidelines was 10CFR50, Appendix A, GDC 19. The following additional radiation limit I guidelines were used to evaluate occupancy and accessibility of plant vital areas 4 e,vt 44ctSJ f3 M Radiation Exposure Guidelines Occupancy Dose Ob'iective Continuous <5 Rem for duration Infrequent I5 Rem for all activities Accessway 710 Rem /hr These dose objectives are for personnel access only. W Thedose[receivedbypersonnelinvitalareasof continuous occupancy should be <15 mrem / W averace over p f 30 days). Thedosefse44rfortheseareasjie, determined-using the control room occupancy factors contained in SRP 6.4, as discussed in NUREG-0737, i e., 1.0 for 0-1 day; 0.6 for 1-4 days; and 0.4 for over 4 days. W W OL' f'S Thedosereceivedbypers[isdeterminedbytakinginto nnel in an infrequent I._ - - - occupancy of vital areass account the frequency and duration of the activities anticipated for that area, and is consistent with i GDC 19 limits. Average area dose rates are used to determine personnel exposure, although local hot spots may exist. .,en eea -egae e se ea w eume _m oee e e y ee e.e w e g w e=W e e,eee ff. ddUtdb Gf /%2SC... ld/w kN6+11 J. 5vit&notettkl fdb bCdY$h W whmd+t7".

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l.yy,g, y_ wyg,.. n y s ~ e-- 'I [ !! ?@f LasrsARb#[~hg%ti_Etri= FvuA na l ya.7 m M 'I.b d. gases could escape from the plant structures and be drawn into j F-a the supply air' intake. l There are four independent monitors, separated in accordance with IEEE 279-1971, that monitor air inside the control room intake duct. These inline monitors respond to the gross radioactivity I in the vicinity of the detectors. Each monitor provides three alarm conditions: low, high, and high-high. The low and high alarms trip the control room annunciator. The high-high alarm trips the control room fresh air isolation valves and starts the control room emergency fresh air supply, which provides for the filtration of the incoming air through HEPA/ charcoal filters. The trip of either monitor A or B shuts off the control room fresh air supply, and the trip of either monitors A and C, or B and D, starts the control room emergency fresh air supply. (See Section 6.4 for a more detailed discussion of control room isolation on detection of high radiation.) 11.5.2.1.5 Control Room Emergency Fresh Air Radiation Monitors This monitoring system is actuated by the tripping of the control room fresh air supply isolation valves. Par'ticulate and iodine 1 radioactive isotopes are removed by the HEPA and charcoal filters. Radioactive noble gas concentration is measured and continuously recorded. These inline monitors detect gross k radiation only. Two monitors, separated in accordance with IEEE 279-1971, monitor sample air from the control room emergency fresh air duct. Each monitor provides three alarm conditions: low, high, and high-high. These alarms trip annunciators in the control room. All requirements of this system are identical with those of the control room air supply radiation monitors except that no provisions for valve closure are made. 11.5.2.1.6 Primary Containment Post-LOCA Radiation Monitor This monitoring system is comprised of four ion chamber sensors for the primary containment in the event of a loss-of-coolant accident (LOCA). After such a postulated accident the monitoring system measures the gross radioactivity present in the containment atmosphere. This information is transmitted to control room personnel to pgovide them with a basis for making safety-related decisions. d<f:f. trip

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~.=.0 0 SHIELDING AND RADIATION i

l I ZONING DRAWING, REACTOR m,*. "~ :r.*,,. 2 ENCLOSURE PLAN AT EL 253*.0" l c. b 1. Fa. 8- 09 l I FIGURE 12 3-16 i

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W %s ?> lk 1 LGS FSAR 3. High-and low-volume portable samplers capable of attaching filters and charcoal cartridges for particulate and iodine monitoring. The ventilation system monitors are located at positions which provide representative air concentrations and a rapid indication of abnormal conditions. Those systems which require HEPA filtration have monitors upstream of the filters. Both the inline Geiger-Muller tube and beta scintillator, and offline particulate, iodine, and nob gas monitoring configurations are utilized. Readout and annu tion are provided in the main control room and/or radwaste ontrol room. Emergency de power is provided in the event of loss of offsite power. The detectors are calibrated routinely and after any maintenance work is performed on the detector. Continuous air monitors (CAMS) are located in freely accessible areas where airborne radioactivity is most likely to exist. These CAMS are mobile and can be moved from area to area as deemed necessary by plant conditions or maintenance operations. CAMS incorporate either fixed or movable filters for the collection of particulate activity, which is monitored directly by a detector. Readout is recorded in CPM. The filters can be removed for further analysis using counting room instrumentation. Audible and visual alarms indicate when set point levels have been exceeded. The detectors are calibrated routinely and after any maintenance work is performed on the detector. The CAM's primary function is to indicate trends and sudden changes in airborne activity. Typical locations are solid waste handling areas, spent fuel pool areas, and the reactor operating floor and turbine building. The monitoring system is capable of detecting ten MPC-hours of particulate and iodine radioactivity from compartments which have a possibility of containing airborne radioactivity, and which normally may be occupied by personnel. A flexible hose can be attached to the monitor intake and inserted into a cavity or work area to detect the presence of localized airborne activity. Conformance to Regulatory Guide 8.2 is discussed in Section 12.5.1. The guidance of Regulatory Guide 8.25 will be followed. ...us... .sa--. u-us.-- n.u.- w. .c

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SENIOR HEALTH PHYSICIST p5 %, I?" J> -1) N /U3 NAME: Kenneth H. Taylor II EDUCATION AND TRAINING 1974 to 1976 U. S. Navy Nuclear Propulsion Schools: Engineering Laboratory Technician (13 wks.), Reactor Plant Operator (26 wks.), Nuclear Power (24 wks.) Machinist Mate (16 wks.) 1981 Purdue University, B.S. In Environmental Health 1982 Purdue University, M.S. in Health Physics 1982 Engineer Orientation Training Program - Four week course in the fundamentals of Nuclear Power Plant Operation WORK EXPERIENCE 6/82 to Senior Health Physicist present Limerick Generating Station assist in pre startup development of Health Physics E Chemistry group and radiation exposurt control and measurement program. Participated in Peach Bottom Atomic Power Station HPEC activities, including procedure development, plant operations, and activity evaluations. 1975 to Assistant Laboratory Manager / Senior Laboratory 6/82 Technician - School of Civil Engineering, Purdue University Supervised design, fabrication, operation, calibration and repair of equipment in t he Hydraulics and Systems Engineering Laboratory. Cost estimating and purchasing. Maintained field data collection stations. Developed and installed a computerized remote sensing data collection system. I i 1 Figure 2.1.1(a) s I

~ 1976 - 1978 Safoty Insp2ctor/Engin2sring Labarctory Tcchnician USS Eisenhow2r, United States N:vy Responsible for safety and industrial hygiene related to reactors, propulsion plants and electrical / electronic equipment. Developed control program for benzene products to cfdD" comply with emergency standard. IkN \\# W MO 05d 1 # O# Assisted in development and implementation of safety procedures. Edited weekly and monthly internal saf ety publications. Technical guidance to departmental safety personnel. 1974-1976 Staff Instructor / Engineering Laboratory Technician U. S. Navy Instructed at D1G Reactor Prototype... radiological controls, reactor plant chemistry, plant theory and operation, emergency and casualty procedures. Performed radiation, contamination and airborne radioactivity surveys; exposure monitoring and centrol personnel dosimetry, radioactive waste processing and accountability, decontamination and related functions in support of reactor refueling and plant overhaul. I l Figure 2.1.1(b) l l

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LGS FSAR WV15e d5 Indtarct TABLE 471.6-1 HEALTH PHYSICS AND CHEMISTRY INSTRUMENTATION INSTRUMENT SENSITIVITY / RANGE $d nt 4wW E knc/ 6i$pectropnotometer LyQ, ))p' -+1-am e 1 ppm -o Gas Chromatograph H, - 30 ppm ; 0,.- 20 ppm 7 /- '~~0 Ion Chromatograph 1 ppb /. _o Photoelectric Colorimeter 1 ppm / o Turbidimeter 0.1 JTU 9 0 U.V. Spectrophotometer 1 ppb T.O.C. Analyzer 50 ppb /. O / o Germanium Counting System 10 min. count (nCi): 3 /_ Co**, Cs23*, Cs137, 1131 lT 1x10-4 2nes Noble Gases - 1x10-8 l Proportional Counter Beta Efficiency 10-50% 2 / (E dependent) Whole Body Counter System <1/20 of the International /_ o Committee of Radiation Protection (ICRP) Pub. 30 l Maximum Permissible Body Burden (MPBB) levels. Pgdlot.u rvedd/.tAL l tRortable 'Etfrvey Instruments l d~0 j_ RM-14 with HP 210T 0 - 50 to 50,000 cpm EO S RM-16 with HP 200 100 cpm / to 1H epm RM-16 with RD 17A 0 - 0.1 mr/hr to 1R/hr E-520 with HP 270 0 - 0.2 mr/hr to 2000 mr/hr . a 2 l Teletector 0 - 0.1 mr/hr to 1000 R/hr ~ // l ' E-520 with HP 230 0 - 0.2 mr/hr to 2000 mr/hr { PRM-6 with AC-3 0 - 500 cpm to 500k cpm l PRS-2P - NRD 0 - 0.2 mr/hr to 10 R/hr . A Aj i RO2A 0 - 50 mr/hr to 50 R/hr b l0 l RO7 1 mr/hr to 20,000 R/hr Q o I RCA 0 - smo tmr/k Js_ _acj / ( y / + I\\ Rev. 18, 03/83 kP au wih wr~ A

wrA DRAFT ( t Quantity seeded 94%dm{3[ masse f Bow Later bothrwdfso, Instruments t 30-2 0-5000 mE/hr 25 25 RD-2A 0-50 t/hr 10 10 E-520 0.2-2000 aE/hr '11 11 4 E-140N 0-5 W epa 11 11 HP270 probe .1 mE/hr-10 R/hr 15 15 15 1'5 EP210AL probe 5 5 HF2107 probe Tele tect or 0-1000 E/hr 3 3 I E04-I (6 cables; 0.01 mE/hr-10,000 R/h 10 10 10 10 DA1-6 probe (*3r !*4*I) PRM-6 with AC-3 500-500,000 cpa 2 2 5-5000 mR/hr 4 0 Pys c 2 0 EP230 - 3 in. sphere 307

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Improved Inplant Iodine Instrumentation Under Accident Conditions (III.D.3.3) Each Licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentrations in areas within the facility where plant persocnel may be present during an accident.

RESPONSE

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Sampling methods and procedures will be implemented at Limerick Generating Station which will permit the measurement of in-plant iodine concentrations during accident conditions. A description of this method is as follows: The sampling method uses portable air samplers with a combination particulate filter and fodine sampling cartridge sampling head. l The sampling heads use a glass fiber particulate filter and a CESCO style (2.25" dia. x 1.04" thickness) iodine charcoal cartridge. The cartridge normally used is the CESCO type charcoal cartridge. When long sampling times are required a larger capacity charcoal cartridge is available. During emergency conditions, with high xenon or krypton concentrations potentially present, either a silver zeolite or a silver impregnated silica-gel adsorber cannis-ter will be employed. i Iodine activity on the sample cartridge will be determined by gamma isotopic analysis using a computer based nailti-channel analyzer with high resolu tion intrinsic germanium detectors located in the Limerick Counting Roc. The Counting Room is located in the Radwaste Enclosure at elevation 217'. An assessment of the NUREG-0737 shielding study indicates that the Counting Room dose rates and airborne radioactivity concentrations are low enough to permit sample analysis during accident conditions. k W i Isotopic adl ysis will permit iodine identification in the presence l of xenon and~ krypton. If the analysis of iodine becomes impossible due to interference (high background) from xenon or krypton, then either silver zeolite cartridges will be used, or the charcoal cartridge will be purged with clean bottled nitrogen or breathin g 4,,g g a If the use of silver zeolite d W air to reduce the interference. not sufficiently reduce the xenon or krypton interference, the silver zeolite cartridges will also be purged with clean bottled nitrogen or bottled breathing air available on site. The Health Physics technical staff will be trained in the implemen-tation of this postaccident procedure. GWM/dg/W/6 -}}

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