NRC Generic Letter 1980-11: Difference between revisions
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{{#Wiki_filter:UNITED STATESNUCLEAR REGULATORY COMMISSIONbVASHItNGTON. 0. C. 20555December 22, 1980TO ALL LICENSEES OF OPERATING PLAtNTS AND , , -FOR OPERATING LICENSES A.D -"HOLDERS OF CONSTRUCTION, PERMITS*Gentlemen:Subject: Control of Heavy Loadsin January 1978, the ',RC published NUREG-04l entitled, Progran -orthe Resolution of Generic Issues Related to Nuclear Power 'lants -Report to Concress." As part of this procra,-, the TasL' Action Plan fzrUnresolved Safety Issue Task 'N;c. 4-36, "Control of Heavy Loads NearSpent Fuel," was issued.ý..!e have completed our review of load handlinc operatiors at nuclearDower plants. A report describing the results of this review has beenissued as NUREG-0612, "Zontrol of Heavy Loads at %uclear Power plants -Resolution of TAP A-36." This -eport contains several recornendationsto be implemented by all licensees and applicants to ensure the safehandling of heavy loads.The purpose of this letter is to request that you review your controlsfor the handling of heavy loads to detern"ine the extent to which thecuidelines of Enclosure 1 are cresently satisfied at your facility, and:0 identify the changes and ,o-ifications that would be -equired inorder to fully satisfy these guidelines.To expedite your compliance wit" this request, we have enclosed thefollowing:,UREG-0'12, "Control of Heavy Loads at '-uclear Power Plants'" Enc'esure1).Staff Position -Interim Actions for Control of Heavy Loads (Enclosure2).Request for Additional Infor-,ation on Control of Heavy Loads (EnclosurewVith the exception of licensees for Indian Point 2 and 3, Zion I and 2and Three Mile Island l (7These Ywere previously sent a letter)~1O&7 3.~-*, A' | {{#Wiki_filter:UNITED STATESNUCLEAR REGULATORY COMMISSIONbVASHItNGTON. 0. C. 20555December 22, 1980TO ALL LICENSEES OF OPERATING PLAtNTS AND , , -FOR OPERATING LICENSES A.D -"HOLDERS OF CONSTRUCTION, PERMITS*Gentlemen:Subject: Control of Heavy Loadsin January 1978, the ',RC published NUREG-04l entitled, Progran -orthe Resolution of Generic Issues Related to Nuclear Power 'lants -Report to Concress." As part of this procra,-, the TasL' Action Plan fzrUnresolved Safety Issue Task 'N;c. 4-36, "Control of Heavy Loads NearSpent Fuel," was issued.ý..!e have completed our review of load handlinc operatiors at nuclearDower plants. A report describing the results of this review has beenissued as NUREG-0612, "Zontrol of Heavy Loads at %uclear Power plants -Resolution of TAP A-36." This -eport contains several recornendationsto be implemented by all licensees and applicants to ensure the safehandling of heavy loads.The purpose of this letter is to request that you review your controlsfor the handling of heavy loads to detern"ine the extent to which thecuidelines of Enclosure 1 are cresently satisfied at your facility, and:0 identify the changes and ,o-ifications that would be -equired inorder to fully satisfy these guidelines.To expedite your compliance wit" this request, we have enclosed thefollowing:,UREG-0'12, "Control of Heavy Loads at '-uclear Power Plants'" Enc'esure1).Staff Position -Interim Actions for Control of Heavy Loads (Enclosure2).Request for Additional Infor-,ation on Control of Heavy Loads (EnclosurewVith the exception of licensees for Indian Point 2 and 3, Zion I and 2and Three Mile Island l (7These Ywere previously sent a letter)~1O&7 3.~-*, A' | ||
I WOMPt--d= -WMWWMWWW1MWddý-2 -December 22, 1980You are requested to implement the interim actions described in Enclosure2 as soon as possible but no later than 90 days from the date of thisletter.In order to enable the NRC to determine whether operating licensesshould be modified (10 CFR 50.54(f)), operating reactor licensees arerequested to provide the following:1. Submit a report documenting the results of your review and therequired changes and modifications. This report shouldinclude the information identified in Sections 2.1 through 2.4of Enclosure 3, on how the guidelines of NUREG-0612 will besatisfied. This report should be submitted in two partsaccording to the following schedule:-Submit the Section 2.1 information within sixmonths from the date of this letter.-Submit the Sections 2.2, 2.3 and 2.4 informationwithin nine months.2. Furnish confirmation within six months that imc'zmentation ofthose changes and modi'ications you find are -essary willcommence as soon as o;!ssible without waitlnc -zaff review,so that all such changes, beyond the above in..-i actions,will be completed within two years of submittal of Section 2.4for the above report.Furnish justification within six months for any changes ormodifications that would be required to fully satisfy theguidelines of Enclosure 1 which you believe are not necessary.T-h :riteria in NUREG-0612 are also applicable to applicants for operatingii:e-ses. Such applicants are expected to provide the informationre:ues:ed by item 1 above and to meet the same schedule of implementationas i;icated in 2 above. Any item for which the implementation date isDrio- to the expected date of issuance of an operating license will beccnsicered to be a prerequisite to obtaining that license.F^r !rv date that cannot be met, furnish a proposed revised date,jus:iication for the delay, and any planned compensating safety actions.zur4: the interim.'Zý 3This requcst for information was approved by GAO under a blanketclearance number R0072 which expires November 30, 1983. Commentson burden and duplication may be directed to the U.S. GeneralAccounting Office, Regulatory Reports Review, Room 5106, 441 G Street,N.W., Washington, D.C. 20548. | I WOMPt--d= -WMWWMWWW1MWddý-2 -December 22, 1980You are requested to implement the interim actions described in Enclosure2 as soon as possible but no later than 90 days from the date of thisletter.In order to enable the NRC to determine whether operating licensesshould be modified (10 CFR 50.54(f)), operating reactor licensees arerequested to provide the following:1. Submit a report documenting the results of your review and therequired changes and modifications. This report shouldinclude the information identified in Sections 2.1 through 2.4of Enclosure 3, on how the guidelines of NUREG-0612 will besatisfied. This report should be submitted in two partsaccording to the following schedule:-Submit the Section 2.1 information within sixmonths from the date of this letter.-Submit the Sections 2.2, 2.3 and 2.4 informationwithin nine months.2. Furnish confirmation within six months that imc'zmentation ofthose changes and modi'ications you find are -essary willcommence as soon as o;!ssible without waitlnc -zaff review,so that all such changes, beyond the above in..-i actions,will be completed within two years of submittal of Section 2.4for the above report.Furnish justification within six months for any changes ormodifications that would be required to fully satisfy theguidelines of Enclosure 1 which you believe are not necessary.T-h :riteria in NUREG-0612 are also applicable to applicants for operatingii:e-ses. Such applicants are expected to provide the informationre:ues:ed by item 1 above and to meet the same schedule of implementationas i;icated in 2 above. Any item for which the implementation date isDrio- to the expected date of issuance of an operating license will beccnsicered to be a prerequisite to obtaining that license.F^r !rv date that cannot be met, furnish a proposed revised date,jus:iication for the delay, and any planned compensating safety actions.zur4: the interim.'Zý | ||
3This requcst for information was approved by GAO under a blanketclearance number R0072 which expires November 30, 1983. Commentson burden and duplication may be directed to the U.S. GeneralAccounting Office, Regulatory Reports Review, Room 5106, 441 G Street,N.W., Washington, D.C. 20548.Sincerely,Darre .IG Eisenhut, DirectorDivision o LicensingEnclosures:1. NUREG-06122. Staff Position3. Request for AdditionalIrformationcc: w/o Enclosure (1)Service List.~ .. | |||
ENCLOSURE 2STAFF POSITION -INTERIM ACTIONS'FORCONTROL OF HEAVY LOADS(1) Safe load paths should be defined per the guidelines of Section5.1.1(1) (See Enclosure 1);Procedures should be developed and implementea per the guidelinesof Section 5.1.1(2) (See Enclosure I);(3) -rane operators should be trained, qualified and conduct themselvesper the guidelines of Section 5.1.1(3) (See'Enclosure 1);Cranes should be inspected, tested, and maintained in accordancewith the guidelines of Section 5.1.1(6) (See Enclosure 1), and(5) In addition to the above, special attention should' e given toprocedures, equipment, and personnel for the handling of heavyloads over the core, such as vessel internals or vessel insoectiontools. This special review should include the following for theseloads: (1) review of procedures for installation of rigging orlifting devices and movement of the load to assure that sufficientdetail is provided and that instructions are clear and concise;(2) visual inspections of load bearing components of cranes, slings,and soecial lifting devices to identify flaws or deficiencies thatcoulch lead to failure of the component; (3) appropriate repair andreplacement of defective components; and (4) verify that the craneoperators have been properly trained and are familiar with specificprocedures used in handling these loads, e.g., hand signals, conducto' operations, and content of procedures. | |||
REQUEST FCt ADDITIO NAL I Nr0CR'T ION ýNCO;JROL OF HEAVY L"ýCUS1. INTRODU:TIONVerificatian by the licensee that the risk associated with :oad-.and..ngfailur,.s at nuclear power plants is extrenely low will require a systema:tic e- al,.a-tion of all load-handling syste-s at each site. The following specific infora:1cr.requests have been organized tz suppor: such a syste-atic approach, and provi-e abasis for the staff's review of the licer.see's evaluation. Additionally, they havebeen organized to address separately the two hazards requiring investiga:ia.on I.e..radic-logical consequences of damage to f!:el and unavailability z-nsequences of:a.-.ae to systems). "he following general information is provided to assIstin this evaluation and reduce the need for clari!Ication as to the Int.-.: arn eN':e-,t-ei- res,;Its of this in;uiry.1. Risk reduc:ion. can be demonstrated by either of two prcac:e-:a. The likelihood of failure is -"de extremely low throuv.h handling-system design features (..EG 0612, Sectiono 1.6).b. The cznsecuences of a failure can be so',-.. to beacceD:able (.7-'.ý 1612, Section 5.1, CrIteria .-Y.ýezardless of the a::roacý selected, the ceneeral i'eidelines.N .-_ 0612, Section 5.1.1, should be satisfied to prc,,,de -? ax..-p:actical de-en.se-in-depth..Evaluations concerning radiological consequences or safety, where used, can rely on either the adcption of ;enericanalyses reported in !:*'-rG 0612, requiring only verficaton thatthese generic asst-ptions are valid.for a secific site or erT.-aya site-s.ecific analy,-sis.3. re;ufred for safe shutdown and continued he.a" heat removalare si:e-s.ec iic, a re are no:, there lore. ldenti' -ed In t.is :e:u :.-...div. ual prants sh.. consider sys:e-s and comn..urnets ident:iztecr.. .e;ula:nry Cui.de 1.:9. .osi-.-cr. C.1 (except those orpor:tons cf systems "..a: are reu.;!red solely for (a) .-2enC'v | |||
* re ,:.) os5t-accident cc:nain-ez" heat re.c-.'al, or tc; .os:-azcrie7:cta:na.n.ent a:-:st;here c'eanur.), for evaluation and :.a:the a-;roach taken. this respect Is PsIAmilar to :-a: iie'.n:I:?.e;u"a:orv Zufde T:sItbcn C.2. 7he fact that a I .'.s:.'s:e- n-.v be .reve-.:e. fro- cnerating dur*ing rant Cond+/-::-ns r-q..r~ng "-e ac.L:ua" cr .o:en" ial -fe of s -e of -,ese s :s rec-POOR ORIUI | |||
o.1 zed in this request for i.r -ormation.4. The scope of this systematic review thould include all'heavr loads tcarried in areas where the potential f r non-cz=_ilanme .i:,n the acceytanc crlteria 1(,L-EG 0612,Section 5.-2) ,xirt.s. A s--ary of typ ical loads to be,considere: bas 'been provided in V1UREG 0612, Table 3.1-1.1: is ized that some cranes -wil1 carry additional=iscellaneous loads. some of which are not identifiablein detail in advance., In such cases an evaluation or.,nal-sis the acceptability of the handling.of a range nf loads should be 5. A:t sme sites -loads which must be eval.atee will includelicensed shipping casks -rDvided for the transportation ofirradiated f:uel. solidified radioactive waste. spent resins.or other byt.--oduc: -aterial. licensing tunder 1OCFR7! is nc:evidence that lifting Ldevices for these shippimg casks nee:-the cr!reria specif ied in n-RSC 06127, Sections 5.. 1. ý ), 5. 11(5). 5.-.6(1.). or 5.1.6(3)., as appropriate, and thus doeseim!i-.t-e the need to provide apprcrpriate inf..-a:-concerning these devices. A tabulation ,C(Atachme:.-_ 5) :sprovided to indicate multiple-sITe use of these ship~ing zasks.-he results of the licensee's evaluation.. as reported in response -. thisrecues:. should 7.rcvide +/-nfcrmation sufficient focr the staff to cdndu:t an in-le'enden: revritw to deter--ine that the intent of this effort (+/-.e., the unifcr-Mre-uct:in of the c:ten-ial hazard fro= load-handling-system failures" has beensa.'.is:ed.2. RE FROM T4iE LICENSEE'2. G 1-.-ZAL REOUI.-EEV- ='.-AD' SYSTEMS,FEZ O61.2, Sectiorn 5.1.,. ecentafies several general guidelines rela:ed -.::he desl.;n and qperaticn cf overhead load-4handling systems in zhe arear whnereF=.en-.: fuel is stzrred, in the vicinity of the reactor cvre, and in c.ner areas of:-e n ia-t wher.e a load drop could result in dam.age to f:7 .a-es ."-o. or deca" 'heat removal&. T for-ma tion provided In resnor-se ".D -t e_-? .n: f--he eý.-._nt off ,nnential'ly hazardous lzad-han.-.!ir. .-7 a aF.-'& and the extenmt cir -ofc'--ance to ap;ron.riate load-hand-lir.g -: ",' results cel v'o,- review e.f pla.-: arrang Fn-:iden-tif'. all r'ver.head handling systems fro, which a 1/2afd r q zav result in dim-te to any systcm ref.uired fcr -::an:s.u -'n..¢e recay heat r7e-'val (takig m c rei.t: fr an-... POOR ORIGINAL | |||
imterlocks, te:..niical specifications. .-;ro:ed.:-es,or detailed s:::ru:ura. analysis).2. Ius:ify theex:usi- of any overhead handlt.ng syste. _ fr.the above r-a:eg.ry =v verifying that there is suffic:ent:phvslzal se:ara:iv-. fr=- ar. , oad-i=-act poi:t a-d a8.safe..-rela-ed _-o_ ponen: to pe-.it a deter= --'y inste:-tien that no hea-v.. load drop can result in da-Age to an.'-syste. or copne:.: required for plant shutdown or deta:?heat redval.3. 'ith respect :o the design- and operation. of heav.-load-handl~ig systems 'in the reactor building a-d those load-handling systems iden:ified in 2.1-1, above, ;rcviLde youre-!valuati.o- con er--in. comlian:e with the guidel.nes ofDe-, Section 5.1.1. .-he following specif:c i.-fr--a-tion should be Included i-m your re;ly:a. or ske.:hes sufficient to clearlyIdentify the loca-io-m of safe load paths, spent.fuel. and safe:y-rela:ed equi--pm:.b. A discussi-o of measures taken to emsure thatioad-handle=g cperations remaim within safe loadpat.hs, in:lud--mg ;rocedures, if any, for dev-a:+/-.=zf!.m these paths.c. A tabula:icm of heavy loads to be handled by eachcrane which includes, the load loa-,weight, its designated lifting device, and verifi-cation that the handling of such load is gcvernedby a written procedure containng., as a =i=i==,the infor--ati=- Iden:4fied in XELG 0612, 5ect.ion5.1.1(2).d. Verification that devices identifled in -. -.13-c. Above, cply wit!h the requirements of AM.%- .Nl.6-1978, or ANSI 330.9-1971 as appropriate. For lif:-ing devices u*ere these standards, as supplementedby X-7E D612., Section 5.1.1(4) or 5.1.1(5). are notmet. describe any ;roposed alternatives and demon-strate t.heL- ,;u:vaiency in ter.ms of load-handlingrellabil+/- i1..e. tha ANSI 330.2-1976, ChaTer 2-2. hasbeen invoked ,-ith respect to crane insnec:ic-_ -es:i-g.and zaintenance. '-:ere any exceptio= .s take- -c :tisstandard, suffic:--t should be provtded tcdemrnstrate the e;ulvalency of prpced al&tea-tIves.f. Verificaticn that crane desin co:=ies with the z-..e-lines of SpecIfica-!.o T0 and Chapter :-' of ANLSIi:lu-i.-g .he Of ecý;.valenc-of actual des.E. for Instances %-here s-ecif:ico- ilance wf:-. :hese standards is -not ;rovtded.-3- | |||
£. Excep:,..s, -4.4 an'.. :,a ken t? S -3.- --e .-e t =tC-r :rai , .-,REAC7'R !%G.Z-AEG 0,51:, Stcti=-- .;rz:'-.;-ds gui-tel-'-s :ý-z and cpera:iom cf 1oad-1 indli.g ss:e' s in- the vici-i:. of S-.-.: .e. "reactcr vessel cT _4n strage. .cr.a:ic ;ra rzvi'ded in :: :7_5se:tio-n should demonstrate t:a.a-a: e;ua:e :Peas-es .-.ave t-ee.- :ake- that, in this area, either :-he :ikel:hood c- a load -cT; '. .-s-em: fuel is exTre_=e2- -r that the es:inated scnse-.en.esa"-rop will no: excee4 t.e li--.ts set .y the evaluatzit cr::er:"a , -.*- -Sectiorn 5.1. Criteria 1 Ttrcuarn ::I.1. Iden-tfv b}y name, :-.-?-e. :a;aziry. and e;,i~me---es:~.a:--any. cranes ;hvr:i~aa7," zacable (i.e.., -t :-.Mzveable mechan:jA! st--.;%s, cr :rD:ed-uresloads over soent fuel !- :he st.rare pczil o-7 .-the react:or vessel.-. :.stifv t*.e exclusicr. of any cranes im thýis area :r-- -.abcve -zor :'. ver-.:n_ :that thev are i.ca:a&!e :zCarr.ing heavy :oads or are ?ernanentl:" :revente,=cve=we of hear. loads over s-:red r it: a:-,location ,,9ere, fo1io-ing a.- failu.re. su-h f"oa' r. ointo the reactor -.-essel or s'ent fuel szcraze -...I. e, ifv a--.; crazes :isted -n 2_2-1. atove, "hZ----Du '-.aveevaluated as havi.-_. sufic:ient design feat-.-res :o =&;,e :-.elikelihood of a load drop extrezely smail zcz a-' loads :-!,e car-tried and the basis for this evaluation (:.e.,co=llance %-+/-:n NtUR-G S. Section 5.1i.1 or 7ar::. z:-pliance sio-plemenzed --- suitable alter-nati:-e or aT --design features). -or each crane so evaiuated, ;rov:_- -.ete- i.e. | |||
* crane-load-oonl.~ati.o) .ioi s;ecified in A::acm-ent 1.4. For cranes iden:lffed -Z.2-1, a!ývve. nct ca:e-.rizzeo a:::7c-To 2..-s. --enns:-rate that t"Ie criteria .. -Seczin 35.1. are sa:isfied. Conpliance wit" Z7i:.-o--'-be demc-ns~ra'.ed in resoz-=se to Sectic.n-. :,._-.:'ith res;rect to Cr:teria : throu ilM. -Tcvi:e a -of vour evaiazin o-f crame o-pera:ion in t:e zeac:or -and your ' eter--natfc-: f =: 1i-ance. .his r::-.einclude te follo..in- i-afcmaaic.n for eazh :rath:a. ;ýere -eliance is ;'_aced on t:e istalla:.cn an_ | |||
.:rz~::.~r~oksor mechanical sto-s, 4ndlca:e.t~ c:cu~tantunder which these ;otectve ccan Ie rTeM.E- or b:.pas sed and the adninistrative 7r---Cuts in+/-,'oked tC ensure proper authcr aat:cncr. an" related or proposed tecnizal soez-'rizations concer.ning the bypass of such +/-n:erlo:Ks..zre reli..nce 's placed cm the operation of the Stand-or *;as Treat=ent System, discuss present andicr prc-osed:ech-Inical sDeci':catins and a4-nins:rative or ,hvs-.ca:controls 'rcv:iedd to ensure that these assu=;tions re-Main va-i-d.-here reliance 4s placed on other site-s;ecfic :con-siderations (e.g., refueling sequencing), prcvide 7resen:Cr ;rocosed technical specIficaticns, and tiscuss ad=nn:s-:ra::ve r. phvs ical controls provided to ensure the valid-4:y of such ccnsiderat;cns.-. Anal,'rses zer:n.ned to demonstrate wh 'Cri:erla "t :nrcuzh shculi conform to the guidelines c' NT7E, Q1/21.Atoendix A. .nus:ifv a exception taken tc -iese g-. ..el.nes,and srovice the s;.ecif'c information re'uestet d At:aznen.. -.3, or ., as a':ro;riate, for each analysis rerfc.-e=:.". 2 STEC!FIC RE' P.9 S FOR OVERH.EAD EN.LING SYSTD'.S OP-.-K: :N ?:N S:O~qAINlNG RKrQU -FOR P.EACTOP. C DE HY EL-.~OR 5?=L "T POOL COOLI;GRU'EG Section 5.1.5, -rovides guidelines concerming the designan! :.eration of load-handling svste=s in the vicinity of t:ui;nen: or co-7-nenzs re;uired for safe reactor and decay heat removal. inf c._a-:>~. 6rovied in respo.se to this section should be suffi-ien: to :emonst:a:t:'a: adecqate measures have been taken to ensure that in i:.ese areas, efther:h.e likelihood of a load drop which might prevent safe reactor shu:ooc orcontinued decay heat removal is extremely smal:, or that danaee tos cuipment from load drops will be li=ited in order not to resui.: rn toeoss of these safe:y-related functions. Cranes which =ust be evaluate- inSect:cn have betn previously identified in your res;onse *o 2. -., and:7eir :ca=s -in "ocur response to 2.1-3-c..:den:ifK anv cranes listed in 2.1-1, above, which vFu xaveevaluated as having sufficient design features to :.akt !-kelihood of a load drop ex:remely mall for all :oa;s :-e zarried and the basis for this evaluation (.e.. cc=:7.:E=:Tl1ance with %2E..G Ot12, Section or ;ar"+/-al cP-:,iance su ;lenen:ed by suitable alternative or additicna:design features). For each crane so evaluated, ;rovide :,e1cad-hand1'-n2-sV'ster. (i.e.. ticn S;ecilfled in Atta.:hzent 1. | |||
2. For any cranes identified in 2.1-1 no: as single-failure-procf in 2.3-1, a ccmprehensive hazard e-aiua:ronshould be prcvided which includes the follcwing +/-n -::en.a. The ;resentaticn in a =a:rix for--a: of all heavvloads and pc:ential i.-act areas where da-agenigh: occur to safety-related equipoen:. Heav"loads identifica:i4n should include designationand weight or cross-reference to !nfor--aticn ;ro-v!ded in 2.1-3-c. Inpact areas should be i-denti-fied by construction zones and elevations or bysome other =ethcd such that the area canbe located on the plant general arrangementdra'ings. Figure 1 provides a typical matrix.b. For each interaction identified, indicate whichof the load and i--pact area co-bInativns can )ebecause of separation and redundanctof safery-related equip-ment, mechanical sC:,sand/or electrical interlocks, or other st:e-specific considera:ions. Elimina:ion on thebasis cf the aforementicned consideration shouldbe SuFlemen:ed 'y the following specifiL inf=zra-tion:(1) For load/target combinatiors elimina:edbecause of separation and redundancy cfsafety-related equipment, discuss thebasis for determining that load dropsillinot affect continued svs:em v-Qera-tion (i.e., the abilit:' of :he syse-:to perform its safety-related func:ion'.(2) `here mechanical stops or electricalinterlocks are to be provided, presentdetails showing the areas where cranetravel will be prohibited. Addirtonal-i. provide a discussion concerning theprocedures that are to be used forauthorizing the bypassing of interlocksor removable stops, for verifying thatinterlocks are functional prior to craneuse, and for verifying that interlocksare restored to o;erabilitv after opera-:+/-ons which require bypassing have beencompleted.(3) ;;here load/target cot 'inations are eli=-inated on the .%asis of other, si:e-s:ec-4f"c censidera:ions (e.g.. =aintenancesequencing), provide present and/cr ;ro-posed technical specifications and dis-cuss ad--inistrative procedures or phvsi-cal cons:rain:s izvoked to ensure thevalidity of such considerations.--i- c. For interactions not eliminated by the analysis of2.3-2-b. above, identif7 any handling systems forspecific loads which you have evaluated as havingsufficient 2esign features to =ake the likelihoodof a load drop ex:re=e2y small and the basis forthis evaluation (i.e., complete compliance witHNUREG 0612, SectiOn 3.l,, or partial cor=liancesupplemented by suitable alternative or addition-al design features). For each so evaluated, pro-vide the load-handling-system (i.e.. crane-load-combination) information specified in Attacl.nent 1.d. For interactions not eli=inated in 2.3-2-b or 2.3-2-c, above, demonstrate using appropriate analysisthat damage would not preclude operation of suffi-cient equipment to allow the system to perform itssafety function following a load drop (KUM7J 0612.Section 5.1, Criterion IV). For each analysis soconducted, the following information should beprovided:(1) An indication of whether or not, for thespecific load baing investigated, theoverhead crane-handling system is designedand constructed such that the hoistingsystem will retain its load in the eventof seismic accelerations equivalent tothose of a safe shutdown earthquake (SSE).(2) The basis for any exceptions taken to theanalytical guidelines of NU.REG 0612, Ap-pendix A.(3) The information requested in Attachment 4.~ | |||
Nc:Es7 TO FI7CUE iNote 1: Indicate ty st-bols :ýe sarezv-relared e~i-j=ent. The licenseeshould provide a list consistent with the clarifiza:tin przvilejin 1.2-3.Note 2: Fazarl Eii=ination Categoriesa. Crane travel for this area/load combination prohibitedbv electrical interlocks or mezhanical step&s.b. System redundancy and separation precludes loss ofcaaabil iry of syste= to perform its safetv-rela:edfunction following this load drop in this area.c. Si:e-specifi: considerations eli=ina:e the nee- to con-sider !oad/equip=ent co=bination.d. Likelihood of handling syste= failure f:r :his i>ad isextremelv s=a7i (:.e. section 5.1.6 tE3 J'2e. Ana2ysis demcnstrates tha: zrane failu.re an: !oa. dropwill no- :a..age safety-related eqi.ipment. | |||
I KAMJ~ ITypical lo~cd/IujiadI Area MIUrfxClAPIR: (iUr~tl~rl TIM CROOKS 01 N.$J ANDI EtAUIrKYW? Ilteala)Ux'AT I ull10$'irATR nif NuILDIMUIS) CONISEPwhIVIM; TUIl, IIO'Air APIAMS FiAAIrLP: RYA1IIl~ OIIlIIM:. AUXILIARY OUILUJIM.(laMIvraf ARIA ST LUS.SrUCTBUM WIJNKS)FJIVAT 100SAFETY -RAlATEDMW IPHISTIALAI.D MIININATIUUCATNU)SEIL.EATIO(SAFETY- IFATF.DnpI nwjnHA1AND FL.INIEATI441(A? iiAM-, t I- I -I---,alm lwallu lloatme)auto ISes, Itu..,, Ls*ad fdeastiII-Catio abouheld locludsdesisemtime mad veight)%post Fuel CooknuI 10/14 (100 me..)I 4 I I ________-96 -I- -4 4---- ------ a ------------- ----. I | |||
S:%3LE-FALUjRE -PR0OO qANJLING SYSTEMSI. Provide the name of the manufacturer and the design-rated load ,W. If:he =axi=L critical load N'CO), as defined in NrREG 0553, is nc: the sameas the DRL, provide this capacity.2. Provide a detailed evaluation of the overhead handling systen with respectto the features of design, fabrication, inspection, testing, and operationas delineated in NUREG 0554 and supplemented by the identified alternativesspecified in NUREG 0612, Appendix C. This evaluation zust include a point-by-point comparison for each section of NUREC 0554. If the alternativeszf N'REG 0612, Appendix C, are used for certain applications in lieu ofcomplying with the reco.-endation of NTREC 0554, this should be explicitlvstated. If an alternative to any of those contained in N'RE5 D554 or NUREG0612, Appendix C, is proposed, details must be provided on the proposedalternative to demonstrate its equivalency.l/3. respect to the seismic analysis employed to demonstrate :hat the over-head handling system can retain the load during a seismic event equal to asafe shutdown earthquake. provide a description of the method of analysis,:he assumptions used, and the mathematical model evaluated in the analvsis..he description of assumptions should include :he basis for selection of:troliley an. load position.A. Provide an evaluation of the lifting devices for each single-failure-proofhandling system with respect to the guidelines of XTREG 0612, Section 5.1.6.5. Provide an evaluation of the interfacing lift points with respect to theguidelines of N12ECEG 0612, Section 5.1.6.1/ if the crane in question nas previously been approved by tne staff as satisfyingVREG 0554, Reg. Guide 1.104, or Part 3 to 2T0-AS09-1, please reference theaate of t-e staff's safety evaluation report or approval letter in liew ;fproviding the information requested by item 2. | |||
= | ý = -: : .. ...... ...--... .. .. ... _ _ -1 --, -.... T .... .... I ... , _ _-- ' Ii.... .......... ...... .. .. .....Ni ....l I I ' MA'z.YS:5 OF R:ZLoG:2AL PELEzEEThe f --ving nr;rI.a:ior. sh..d be ;rov"ied fcr an analvsis ccn_'du:ed toCEzmnstra:e cot7fianze with Cri:er4on 1 of N .REZ; 0612, Sec:ion 5.1....I.IAL CO....SiASSL.7TcN$a. ldentif-.* the time after shutdown, the number of fuelassemblies damaged. and the assumed curation of radio-lcgical release associated with eacn accident analvzed.b. NL2EG 0612, Table 2.1-2, prcvides the asaumptions usedto arrive at generic conclusions concerning radiolcgicaldose consezuences. To rely on the radlological doseanalysis of NUREC 0612, the licensee should ".'erifv Thatthese assunD:iors are zonservat4.,i 1:1:h regar2 :t thePlant/siTe evaluated. if the assume:ions are noc con-seetva4-e for the pe: ific 7lant, or if a =cre site-specific analysis is required, the licensee shou! 2identifv plant-s-ecific assumptions used in place cfthose tab4lazed.c. Identify and provide the basis (e.e., VSNRC RegulatoryGuide 1.25) for any assu=ptions employed in site-specificanalyses not identified in KUREG 0612, Table 2.1-2.d. Dose calculations based on the termination or mi:iga:ionof radtolouical releases should be supported bv inf:--ra-tion sufficient tc demonstrate both that the ti=e ýelavassuzed is conserva:ive and that the syste-_ p:cvided toaccomplish such termnna:ion or mitigation will :erformits safety function jpon demand (i.e.. tne system meetsthe criteria for an Engineered Safety Feature). Specificinfor-mation so proviced should include the follow:ng:(1) Details concerning the loca:ion of accidentsensors, parameters zonitcred and the valuescf these parameters at which a safety signalwill be initiated, sys:e= response t Ime(Including valve-operation time), and thetotal ti=e required to auto=atically shiftfro= nor--al operation to isolation or filtra-tion following an accident.(2) A description of the ins:rumenta:ion and con-trols associated with the Engineered Safer:Feature which includes Infcrmation sufficien:to dencnstrate :h;.z the re;jire=ents (Secticn 4)of 1EEE 279-1971, "Criteria for ProtectionSyste=s for Nuclear Power Generating Stations,"are satisfIed.7T.. | ||
(3) A description of any Engineered SafetyFeature filter system which includes infor-mation sufficient to demonstrate compliancewith the guidelines of USNRC RegulatoryGuide 1.52, "Design, Testing, and MaintenanceCriteria for Engineered Safety Feature Atmos-phere Cleanup System Air Filtration andAbsorption Units of Light-Water-CooledNuclear Power Plants."(4) A discussion of any initial conditions,e.g., manual valves lo:ked shut, containmentairlocks or equipment hatches shut) necessaryto ensure that releases will be terminated ormitigated upon Engineered Safety Featureactuation and the measures employed (i.e., Tech-nical Specification and administrative controls)to ensure that these initial conditions aresatisfied and that Engineered Safety Featuresystems are operable prior to the load lift.2. METHOD OF ANALYSISDiscuss the method of analysis used to demonstrate that post-accident dosewill be well within 10CFM00 limits. In presenting methodology used indetermining the radiological consequences, the following informaticn shouldbe provided.a. A description of the mathematical or physical modelemployed.b. An identification and sumary of any computer programused in this analysis.c. The consideration of uncertainties in calculationalmethods, equipment perfor=-ance, instrumentationresponse characteristics, or other indeterminateeffects taken into account in the evaluation of theresults.3. CONCLUSIONProvide an evaluation comparine the results of the analysis to Cri:ericn io,7 'REC 0612, Section 5.1. If the postulated heavv-load-dr:-p a:ccen:a.alyzed bounds other -cs:-lated heavy-load drops, a lisL cf these boundedheavy loads.should be provided.* , :71 + .*., Aw. M 5.7--W WUT.'7". -.... L ..7 *N O M41 -..UM.M | |||
Attachment (3)CRITICALITY ANALYSISThe following information should be provided for analysis conducted to demon-strata compliance with Criterion II of NUMEG 0612, Section 5.11. INITIAL CON'DITIONS/ASSLWTIONSThe conclusions of NUR.G 0612, Section 2.2, are based on a particularmodel fuel assembly. If a licensee uses the results of Section 2.2rather than performing an independent neutronics analysis, the assump-tions should be verified to be compatible with plant-specific design.For any analysis conducted, the following assumptions should be providedas a minimum:a. Water/UO2 volume ratiob. The boron concentration for the refueling waterand spent-fuel poolc. The amount of neutron poison in the fueld. Fuel enrichmente. The reactivity insertion value due to crushing ofthe coref. T-he kIfc value allowed by technical specificationsfor t~e core during refueling2. .MTHOD OF ANALYSISProvide the method of analysis used to dezonstrate that accidentaldropping of& heavy load does not result in a configuration of the fuelsuch that keff is larger than 0.95. The discussion of the method ofanalysis should include the following infortation:a. Identification of the computer codes employedb. A discussion of allowances or compensation forcalculation and physical uncertainties3. CONCLUSIONProv.de an evaluation co=paring the results of the analysis to Criterion IIof LKUREG 0612, Section 5.1. If the postulated heavy-load-drop accident3-1MM-, | |||
bounds other postulated heavy-load drops, a list of these bo'--.ded heavyloads should be ;rovided. | |||
&# | |||
......... ......, .- ..... ..... -.. ... ........II .......... ... ....... .-i i -Attacnr-nent (4)ANALYSIS OF PLANT STRUCTURESThe following infor--ation should be provided for analyses conducted to demon-strate co=pliance with Criteria !I! and IV of N'*REC 0612, Section 5.1.I. !NITILAL CONDITIONS/ASSL 57TIONSDiscuss the assumptions used in the analysis, including:a. Weight of heavy loadb. Impact area of loadc. Drop heightd. Drop locatione. regarding credit taken in the analysis fcrthe action of i=pact limitersf. Thickness of-walls or floor slabs impactedg. Assutotions regarding drag forces caused by theenvironmenth. Load combinations consideredi. Material properties of steel and concrete2. ~~A.0' OF ANALYSIS?rcvide the method cf analysis used to demonstrate that sufficien: load-carrying capability exists within the wall(s) or floor identifyany co=puter codes enployed, and provide a description of their capa.ilities.If test data was employed, provide it and describe its applicabi2ity.3. CONCLUSION?rovide an evaluation comparing the results cf this analysis with CriteriaIII and iV of KnEC 0612, Section 5.1. '*here safe-shutdow- eq-Ipoent hasa ceiling or wall separating it fro= an overhead handling syste=, pr:':idean evalua:ica to demonstrate that postulated load drops do not ;tne:ratethe ceiling or cause secondary missiles that could prever: a s.: ;te= from perfor'.ing its safety function.TM OT | |||
J_I of 8SHIELDED SHIPP-I.,5 CASKSFOR NK"LEAR P0PERCERT, PLA.NT$SI -Fuel (Npwi and Spent)CEtR. 4986545058C5593159386078b 0 662736375LiL-1, 2, 3, 1.ICC, 1. 2, 3Va.-denburghXTS Model 10092 7 Cl927CI348 (Series)?3_27l'.APY LICENS-ESGeneral Elec:ric Co.Westinghouse ElectricChez-Nuclear Systems,Inc.Nuclear Fuel ServicesCobustlon Engineer-Uabcock & Wilcox Co.Che-Nuclear Syste=s,Inc.Westinghouse Elec:ticCo.Nuclear Fuel Services,inc.Cenera1 Electic Co.)7" Industries, In:.General Electric Co.CROSS LOT INUS. AP3.70,000126,20048,000620070006940450067,05045,00050,000ayE?, DLCAPC, CPL, DLP, DPC,TPL, FPC, JC?, %%?P,yE?ac, PCEPECAPL:PC. F?CVt?APO, 3EC,14'L, TPC,nX''A ?C.Pry ,, TVA,AP.-. CPC,".TC, '??.BCE, XYC, .?-.l.n'^ .E,6, 4 01 Su;er Tiger669r,F? ,SCE,DCJC?.NS?,ZL?,900190.1IF 300N- LO12Cr-160,)47,50023,000CPL."ArC,CpC.Cpc.E c,C.'E%.S?,C?!.,P7:.~f-' A.- -- | |||
7-772 of 6SHIELDED SiP;!NuG CASPS CERTIFICATEDF N.LEAR POWER PLANTSII -'asteCERT. mOw_'/'PRIYJ'.AY LIC-NSrE-GROSS LOT INLES. (APWFoX.)S %CO-'DY LICE--NSE'5026BC-48-220605861446244B3-16144C'he.=-Nuiclear Syste~s,Inc.Nuclear E--Sneering Co.Nuclear Lngineering Co.Chem-N-jclear Syste=s,Inc.71,00030,00042,000AC,CYA,FC,? SA?L.MIC,AC,C? C,I:- C,Any,A.-C,FPLI%---_2BEC,DC,JC?,CPC,N?,7EC,Dn?,CPL,F? C,CL., Cv---,DLC, -PL,NPP, VZ.,DL?, IEl,NS?, ?CE,VP?C?.L,K.S?,GC,NSP,CEC,* .4.9'v~ r.3 C?,?EC,624446,0006272 d oly Pan:her6568 LL-60-1506574 le 200Nuclear Engineeriog Cc.Ternessee Valley Auth.Hitt=an Nuclear andDevelopnent Corp.Che=-N\clear Sys:ezs,Inc.61-00AFL., CC, DL?, KC73,00047,0006601LL-50-10070,000A? L,DLP,AC,CEC,F7rL,vz?iDLC,E C,EEC.CC,1-z,.Y A:--F- ,66` 9V!2 SuerTigerNuclear Engineering Cc.Tenessee Valley A-:h.45,00051,0036712 S-33-:50C: .-tnt.... ,.~------ | |||
3 of 6SMIELDED SHIPPING CASKS CERTIFICATEDFOR NUCLEAR POWER PLANTSII -WasteCz-n-& .Xc:)FLPRIXARY .LIC-SEECT.OS S LOT INLBS. (APPROX.)6744Poly TigerNuclear Engineering Co.Nuclear Enginaering Co.6771 SN-I9074A?-1009079 M;-100 Ser. 290o0 \-60090S6 Y. 100 Set. IHittman .Nuclear andDevelopment Corp.RPittman Nuclear andDevelopment Corp.Pittman Nuclear andDevelopment Corp.Pit:man Nuclear andDevelopment Corp.Pitt.an Nuclear andDevelopment Corp.HiBtman Nuclear andDevelopment Corp,.Cheer-Nuclear Systems,lnc.35,00060,00028,00098,00042,00046,000SECONARY LICI:ýSEtAPL, BEC, CC, DL?,)IC, NP?, Sm., vr?APL, CPC, DLP, '%P,SY., VEtDLCAPL,DLY,MCE,AL,N??,EGE,ILE,N?P,NCE,NNE,CEC, Cw-,JCP, -.%A,P ZcCEC,JC? IPEC,XY.'A,DL.-'nA.YACVt. f:RG E,9CS9 Y-\-IOCS9092 EN-3009093 L-40036,50043,00043,00056,500BGE, C'.E, CiC, :.=:,JCP, n'A, P??, ?ECMYAHYA9094CNSI-14-195-HAPC.M*E,DC,JCP,INS?PCC'VE?AMC,DC,JCP,APL,CYA,F'?L,O?,A?L,VPE,PEG,CE-,T.C,PEG,CL,T?C,VPEPCPL,CPC.,C?C,FEC,TV A,C".-,Op: .9096CNSI-2.1-300Chem-Nuclear Systems,Inc.57,450*sc.( At~a'.&Zlf 1-4s | |||
4 of ESHIELDED SH;PPINS CAS.S CERTIFICATEDFOR NUCLEAR POWER PLANTS!I -WasteGRDSS LOT It.LBS. (APPROX.)CERT. 'ODL9105R:-LWaste CR.I9105 AL-33-90PRAk Y.- LICENSEEChie-Nuclear Systems,Inc.Che--Nuclear Systems,Inc.Cbe=-Nuclear Syste-s,Inc.Chem-Nuclear Systems.Inc.Che-Nuclear Syste-s,Inc.58,40041.300SECOV'DA.RY L:Cý;SEi *APC,A?C,DPC,h?P,AC,DPC,Vt?APC,CYA,C?C,91i39113C1,6-80A7-10051,500CL, DC, FL,G?C, JCP,VuCL,D.?C,NST,C'- ,FPC,F?C,C?-,CTC.pzC,7000F? C,V!N-2_,9122 ia--.5061,0003!CSei a :-e. | |||
1~Attach.-.en (5)5 of 6SI"IELD7ED SHIPPING CASKS CERTIFICATEDFCk NUCLEAvR POUER PLAIJSIII -?yprodejctsCTR-1 .PRLV.RY LICENSEEEGROSS LOT INLES. (APPROX.)SECON&DRY LICENSEEPLC5971GE-2005980 .--60010,00018,50030,00026,000P.;E, NS?62759081LL-26-4CNS-1600Che--Nuclear Systems,Inc.Chem-Nuclear Syste=m,Inc.APC, CPL. DPC, FPL,FPC, N??, VE?APC. BGE, CL, DPC,FM1., FPC, GFC, NSP,TVA, VE?Seeof | |||
~JAttaChment (5)L4CE':tE AStEVIA7IONS 6 of 6APC Alabama Power CompanyAPL Arkansas Power and Light CompanyBEC Boston Edison CompanyBGE Baltimore Gas and Electric CompanyCEC Consolidated Edison CompanyCPC Consu=ers Power CompanyCL Carolina Power and Light CompanyC1W Co-onwealth Edison CompanyCYA Connecticut Yankee Atomic Power CompanyDLC Duquesne Light CotpanyDLP Dairyland Power CooperativeDPC Duke Pover CompanyFPC Florida Power CorporationFPL Florida Power and Light CompanyGPC Georgia Power Co=panyIEL Iowa Electric Light and Power CompanyL Indiana and Michigan Electric CompanyJCP Jersey Central Power and Light-Company?CC Metropolitan Edison CompanyMYA Maine Yankee Atomic Power Companyh-vT Niagara Mohawk Power CorporationIN Northeast Nuclear Energy CompanyNP Nebraska Public Power CorporationNSP Northern States Power CompanyOPP Omaha Public Power DistrictPEI Philadelphia Electric CompanyPEG Public Service Electric and Gas CompanyPCC Portland General Electric CompanyPY Power Authority of the State of New YorkRGC Rochester Gas and Electric CorporationSnJ Sacramento Munici;al Utilities CorporactonTEC Toledo Edison Cor.,anyTVA Tennessee Valley AuthorityV`P Vir;inra Eletric and Power Co=panyVyC Vercnt Yankee Nuclear Power Cor4rastionYAC Yankee Atotic Zlectirc Co=parny~ Powe~r Cor.-any-~ iscS.:- p ~ cce Ccrporaticn | |||
}} | |||
~JAttaChment (5)L4CE':tE AStEVIA7IONS 6 of 6APC Alabama Power CompanyAPL Arkansas Power and Light CompanyBEC Boston Edison CompanyBGE Baltimore Gas and Electric CompanyCEC Consolidated Edison CompanyCPC Consu=ers Power CompanyCL Carolina Power and Light CompanyC1W Co-onwealth Edison CompanyCYA Connecticut Yankee Atomic Power CompanyDLC Duquesne Light CotpanyDLP Dairyland Power CooperativeDPC Duke Pover CompanyFPC Florida Power CorporationFPL Florida Power and Light CompanyGPC Georgia Power Co=panyIEL Iowa Electric Light and Power CompanyL Indiana and Michigan Electric CompanyJCP Jersey Central Power and Light-Company?CC Metropolitan Edison CompanyMYA Maine Yankee Atomic Power Companyh-vT Niagara Mohawk Power CorporationIN Northeast Nuclear Energy CompanyNP Nebraska Public Power CorporationNSP Northern States Power CompanyOPP Omaha Public Power DistrictPEI Philadelphia Electric CompanyPEG Public Service Electric and Gas CompanyPCC Portland General Electric CompanyPY Power Authority of the State of New YorkRGC Rochester Gas and Electric CorporationSnJ Sacramento Munici;al Utilities CorporactonTEC Toledo Edison Cor.,anyTVA Tennessee Valley AuthorityV`P Vir;inra Eletric and Power Co=panyVyC Vercnt Yankee Nuclear Power Cor4rastionYAC Yankee Atotic Zlectirc Co=parny~ Powe~r Cor.-any-~ iscS.:- p ~ cce Ccrporaticn}} | |||
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Revision as of 17:22, 6 April 2018
UNITED STATESNUCLEAR REGULATORY COMMISSIONbVASHItNGTON. 0. C. 20555December 22, 1980TO ALL LICENSEES OF OPERATING PLAtNTS AND , , -FOR OPERATING LICENSES A.D -"HOLDERS OF CONSTRUCTION, PERMITS*Gentlemen:Subject: Control of Heavy Loadsin January 1978, the ',RC published NUREG-04l entitled, Progran -orthe Resolution of Generic Issues Related to Nuclear Power 'lants -Report to Concress." As part of this procra,-, the TasL' Action Plan fzrUnresolved Safety Issue Task 'N;c. 4-36, "Control of Heavy Loads NearSpent Fuel," was issued.ý..!e have completed our review of load handlinc operatiors at nuclearDower plants. A report describing the results of this review has beenissued as NUREG-0612, "Zontrol of Heavy Loads at %uclear Power plants -Resolution of TAP A-36." This -eport contains several recornendationsto be implemented by all licensees and applicants to ensure the safehandling of heavy loads.The purpose of this letter is to request that you review your controlsfor the handling of heavy loads to detern"ine the extent to which thecuidelines of Enclosure 1 are cresently satisfied at your facility, and:0 identify the changes and ,o-ifications that would be -equired inorder to fully satisfy these guidelines.To expedite your compliance wit" this request, we have enclosed thefollowing:,UREG-0'12, "Control of Heavy Loads at '-uclear Power Plants'" Enc'esure1).Staff Position -Interim Actions for Control of Heavy Loads (Enclosure2).Request for Additional Infor-,ation on Control of Heavy Loads (EnclosurewVith the exception of licensees for Indian Point 2 and 3, Zion I and 2and Three Mile Island l (7These Ywere previously sent a letter)~1O&7 3.~-*, A'
I WOMPt--d= -WMWWMWWW1MWddý-2 -December 22, 1980You are requested to implement the interim actions described in Enclosure2 as soon as possible but no later than 90 days from the date of thisletter.In order to enable the NRC to determine whether operating licensesshould be modified (10 CFR 50.54(f)), operating reactor licensees arerequested to provide the following:1. Submit a report documenting the results of your review and therequired changes and modifications. This report shouldinclude the information identified in Sections 2.1 through 2.4of Enclosure 3, on how the guidelines of NUREG-0612 will besatisfied. This report should be submitted in two partsaccording to the following schedule:-Submit the Section 2.1 information within sixmonths from the date of this letter.-Submit the Sections 2.2, 2.3 and 2.4 informationwithin nine months.2. Furnish confirmation within six months that imc'zmentation ofthose changes and modi'ications you find are -essary willcommence as soon as o;!ssible without waitlnc -zaff review,so that all such changes, beyond the above in..-i actions,will be completed within two years of submittal of Section 2.4for the above report.Furnish justification within six months for any changes ormodifications that would be required to fully satisfy theguidelines of Enclosure 1 which you believe are not necessary.T-h :riteria in NUREG-0612 are also applicable to applicants for operatingii:e-ses. Such applicants are expected to provide the informationre:ues:ed by item 1 above and to meet the same schedule of implementationas i;icated in 2 above. Any item for which the implementation date isDrio- to the expected date of issuance of an operating license will beccnsicered to be a prerequisite to obtaining that license.F^r !rv date that cannot be met, furnish a proposed revised date,jus:iication for the delay, and any planned compensating safety actions.zur4: the interim.'Zý
3This requcst for information was approved by GAO under a blanketclearance number R0072 which expires November 30, 1983. Commentson burden and duplication may be directed to the U.S. GeneralAccounting Office, Regulatory Reports Review, Room 5106, 441 G Street,N.W., Washington, D.C. 20548.Sincerely,Darre .IG Eisenhut, DirectorDivision o LicensingEnclosures:1. NUREG-06122. Staff Position3. Request for AdditionalIrformationcc: w/o Enclosure (1)Service List.~ ..
ENCLOSURE 2STAFF POSITION -INTERIM ACTIONS'FORCONTROL OF HEAVY LOADS(1) Safe load paths should be defined per the guidelines of Section5.1.1(1) (See Enclosure 1);Procedures should be developed and implementea per the guidelinesof Section 5.1.1(2) (See Enclosure I);(3) -rane operators should be trained, qualified and conduct themselvesper the guidelines of Section 5.1.1(3) (See'Enclosure 1);Cranes should be inspected, tested, and maintained in accordancewith the guidelines of Section 5.1.1(6) (See Enclosure 1), and(5) In addition to the above, special attention should' e given toprocedures, equipment, and personnel for the handling of heavyloads over the core, such as vessel internals or vessel insoectiontools. This special review should include the following for theseloads: (1) review of procedures for installation of rigging orlifting devices and movement of the load to assure that sufficientdetail is provided and that instructions are clear and concise;(2) visual inspections of load bearing components of cranes, slings,and soecial lifting devices to identify flaws or deficiencies thatcoulch lead to failure of the component; (3) appropriate repair andreplacement of defective components; and (4) verify that the craneoperators have been properly trained and are familiar with specificprocedures used in handling these loads, e.g., hand signals, conducto' operations, and content of procedures.
REQUEST FCt ADDITIO NAL I Nr0CR'T ION ýNCO;JROL OF HEAVY L"ýCUS1. INTRODU:TIONVerificatian by the licensee that the risk associated with :oad-.and..ngfailur,.s at nuclear power plants is extrenely low will require a systema:tic e- al,.a-tion of all load-handling syste-s at each site. The following specific infora:1cr.requests have been organized tz suppor: such a syste-atic approach, and provi-e abasis for the staff's review of the licer.see's evaluation. Additionally, they havebeen organized to address separately the two hazards requiring investiga:ia.on I.e..radic-logical consequences of damage to f!:el and unavailability z-nsequences of:a.-.ae to systems). "he following general information is provided to assIstin this evaluation and reduce the need for clari!Ication as to the Int.-.: arn eN':e-,t-ei- res,;Its of this in;uiry.1. Risk reduc:ion. can be demonstrated by either of two prcac:e-:a. The likelihood of failure is -"de extremely low throuv.h handling-system design features (..EG 0612, Sectiono 1.6).b. The cznsecuences of a failure can be so',-.. to beacceD:able (.7-'.ý 1612, Section 5.1, CrIteria .-Y.ýezardless of the a::roacý selected, the ceneeral i'eidelines.N .-_ 0612, Section 5.1.1, should be satisfied to prc,,,de -? ax..-p:actical de-en.se-in-depth..Evaluations concerning radiological consequences or safety, where used, can rely on either the adcption of ;enericanalyses reported in !:*'-rG 0612, requiring only verficaton thatthese generic asst-ptions are valid.for a secific site or erT.-aya site-s.ecific analy,-sis.3. re;ufred for safe shutdown and continued he.a" heat removalare si:e-s.ec iic, a re are no:, there lore. ldenti' -ed In t.is :e:u :.-...div. ual prants sh.. consider sys:e-s and comn..urnets ident:iztecr.. .e;ula:nry Cui.de 1.:9. .osi-.-cr. C.1 (except those orpor:tons cf systems "..a: are reu.;!red solely for (a) .-2enC'v
- re ,:.) os5t-accident cc:nain-ez" heat re.c-.'al, or tc; .os:-azcrie7:cta:na.n.ent a:-:st;here c'eanur.), for evaluation and :.a:the a-;roach taken. this respect Is PsIAmilar to :-a: iie'.n:I:?.e;u"a:orv Zufde T:sItbcn C.2. 7he fact that a I .'.s:.'s:e- n-.v be .reve-.:e. fro- cnerating dur*ing rant Cond+/-::-ns r-q..r~ng "-e ac.L:ua" cr .o:en" ial -fe of s -e of -,ese s :s rec-POOR ORIUI
o.1 zed in this request for i.r -ormation.4. The scope of this systematic review thould include all'heavr loads tcarried in areas where the potential f r non-cz=_ilanme .i:,n the acceytanc crlteria 1(,L-EG 0612,Section 5.-2) ,xirt.s. A s--ary of typ ical loads to be,considere: bas 'been provided in V1UREG 0612, Table 3.1-1.1: is ized that some cranes -wil1 carry additional=iscellaneous loads. some of which are not identifiablein detail in advance., In such cases an evaluation or.,nal-sis the acceptability of the handling.of a range nf loads should be 5. A:t sme sites -loads which must be eval.atee will includelicensed shipping casks -rDvided for the transportation ofirradiated f:uel. solidified radioactive waste. spent resins.or other byt.--oduc: -aterial. licensing tunder 1OCFR7! is nc:evidence that lifting Ldevices for these shippimg casks nee:-the cr!reria specif ied in n-RSC 06127, Sections 5.. 1. ý ), 5. 11(5). 5.-.6(1.). or 5.1.6(3)., as appropriate, and thus doeseim!i-.t-e the need to provide apprcrpriate inf..-a:-concerning these devices. A tabulation ,C(Atachme:.-_ 5) :sprovided to indicate multiple-sITe use of these ship~ing zasks.-he results of the licensee's evaluation.. as reported in response -. thisrecues:. should 7.rcvide +/-nfcrmation sufficient focr the staff to cdndu:t an in-le'enden: revritw to deter--ine that the intent of this effort (+/-.e., the unifcr-Mre-uct:in of the c:ten-ial hazard fro= load-handling-system failures" has beensa.'.is:ed.2. RE FROM T4iE LICENSEE'2. G 1-.-ZAL REOUI.-EEV- ='.-AD' SYSTEMS,FEZ O61.2, Sectiorn 5.1.,. ecentafies several general guidelines rela:ed -.::he desl.;n and qperaticn cf overhead load-4handling systems in zhe arear whnereF=.en-.: fuel is stzrred, in the vicinity of the reactor cvre, and in c.ner areas of:-e n ia-t wher.e a load drop could result in dam.age to f:7 .a-es ."-o. or deca" 'heat removal&. T for-ma tion provided In resnor-se ".D -t e_-? .n: f--he eý.-._nt off ,nnential'ly hazardous lzad-han.-.!ir. .-7 a aF.-'& and the extenmt cir -ofc'--ance to ap;ron.riate load-hand-lir.g -: ",' results cel v'o,- review e.f pla.-: arrang Fn-:iden-tif'. all r'ver.head handling systems fro, which a 1/2afd r q zav result in dim-te to any systcm ref.uired fcr -::an:s.u -'n..¢e recay heat r7e-'val (takig m c rei.t: fr an-... POOR ORIGINAL
imterlocks, te:..niical specifications. .-;ro:ed.:-es,or detailed s:::ru:ura. analysis).2. Ius:ify theex:usi- of any overhead handlt.ng syste. _ fr.the above r-a:eg.ry =v verifying that there is suffic:ent:phvslzal se:ara:iv-. fr=- ar. , oad-i=-act poi:t a-d a8.safe..-rela-ed _-o_ ponen: to pe-.it a deter= --'y inste:-tien that no hea-v.. load drop can result in da-Age to an.'-syste. or copne:.: required for plant shutdown or deta:?heat redval.3. 'ith respect :o the design- and operation. of heav.-load-handl~ig systems 'in the reactor building a-d those load-handling systems iden:ified in 2.1-1, above, ;rcviLde youre-!valuati.o- con er--in. comlian:e with the guidel.nes ofDe-, Section 5.1.1. .-he following specif:c i.-fr--a-tion should be Included i-m your re;ly:a. or ske.:hes sufficient to clearlyIdentify the loca-io-m of safe load paths, spent.fuel. and safe:y-rela:ed equi--pm:.b. A discussi-o of measures taken to emsure thatioad-handle=g cperations remaim within safe loadpat.hs, in:lud--mg ;rocedures, if any, for dev-a:+/-.=zf!.m these paths.c. A tabula:icm of heavy loads to be handled by eachcrane which includes, the load loa-,weight, its designated lifting device, and verifi-cation that the handling of such load is gcvernedby a written procedure containng., as a =i=i==,the infor--ati=- Iden:4fied in XELG 0612, 5ect.ion5.1.1(2).d. Verification that devices identifled in -. -.13-c. Above, cply wit!h the requirements of AM.%- .Nl.6-1978, or ANSI 330.9-1971 as appropriate. For lif:-ing devices u*ere these standards, as supplementedby X-7E D612., Section 5.1.1(4) or 5.1.1(5). are notmet. describe any ;roposed alternatives and demon-strate t.heL- ,;u:vaiency in ter.ms of load-handlingrellabil+/- i1..e. tha ANSI 330.2-1976, ChaTer 2-2. hasbeen invoked ,-ith respect to crane insnec:ic-_ -es:i-g.and zaintenance. '-:ere any exceptio= .s take- -c :tisstandard, suffic:--t should be provtded tcdemrnstrate the e;ulvalency of prpced al&tea-tIves.f. Verificaticn that crane desin co:=ies with the z-..e-lines of SpecIfica-!.o T0 and Chapter :-' of ANLSIi:lu-i.-g .he Of ecý;.valenc-of actual des.E. for Instances %-here s-ecif:ico- ilance wf:-. :hese standards is -not ;rovtded.-3-
£. Excep:,..s, -4.4 an'.. :,a ken t? S -3.- --e .-e t =tC-r :rai , .-,REAC7'R !%G.Z-AEG 0,51:, Stcti=-- .;rz:'-.;-ds gui-tel-'-s :ý-z and cpera:iom cf 1oad-1 indli.g ss:e' s in- the vici-i:. of S-.-.: .e. "reactcr vessel cT _4n strage. .cr.a:ic ;ra rzvi'ded in :: :7_5se:tio-n should demonstrate t:a.a-a: e;ua:e :Peas-es .-.ave t-ee.- :ake- that, in this area, either :-he :ikel:hood c- a load -cT; '. .-s-em: fuel is exTre_=e2- -r that the es:inated scnse-.en.esa"-rop will no: excee4 t.e li--.ts set .y the evaluatzit cr::er:"a , -.*- -Sectiorn 5.1. Criteria 1 Ttrcuarn ::I.1. Iden-tfv b}y name, :-.-?-e. :a;aziry. and e;,i~me---es:~.a:--any. cranes ;hvr:i~aa7," zacable (i.e.., -t :-.Mzveable mechan:jA! st--.;%s, cr :rD:ed-uresloads over soent fuel !- :he st.rare pczil o-7 .-the react:or vessel.-. :.stifv t*.e exclusicr. of any cranes im thýis area :r-- -.abcve -zor :'. ver-.:n_ :that thev are i.ca:a&!e :zCarr.ing heavy :oads or are ?ernanentl:" :revente,=cve=we of hear. loads over s-:red r it: a:-,location ,,9ere, fo1io-ing a.- failu.re. su-h f"oa' r. ointo the reactor -.-essel or s'ent fuel szcraze -...I. e, ifv a--.; crazes :isted -n 2_2-1. atove, "hZ----Du '-.aveevaluated as havi.-_. sufic:ient design feat-.-res :o =&;,e :-.elikelihood of a load drop extrezely smail zcz a-' loads :-!,e car-tried and the basis for this evaluation (:.e.,co=llance %-+/-:n NtUR-G S. Section 5.1i.1 or 7ar::. z:-pliance sio-plemenzed --- suitable alter-nati:-e or aT --design features). -or each crane so evaiuated, ;rov:_- -.ete- i.e.
- crane-load-oonl.~ati.o) .ioi s;ecified in A::acm-ent 1.4. For cranes iden:lffed -Z.2-1, a!ývve. nct ca:e-.rizzeo a:::7c-To 2..-s. --enns:-rate that t"Ie criteria .. -Seczin 35.1. are sa:isfied. Conpliance wit" Z7i:.-o--'-be demc-ns~ra'.ed in resoz-=se to Sectic.n-. :,._-.:'ith res;rect to Cr:teria : throu ilM. -Tcvi:e a -of vour evaiazin o-f crame o-pera:ion in t:e zeac:or -and your ' eter--natfc-: f =: 1i-ance. .his r::-.einclude te follo..in- i-afcmaaic.n for eazh :rath:a. ;ýere -eliance is ;'_aced on t:e istalla:.cn an_
.:rz~::.~r~oksor mechanical sto-s, 4ndlca:e.t~ c:cu~tantunder which these ;otectve ccan Ie rTeM.E- or b:.pas sed and the adninistrative 7r---Cuts in+/-,'oked tC ensure proper authcr aat:cncr. an" related or proposed tecnizal soez-'rizations concer.ning the bypass of such +/-n:erlo:Ks..zre reli..nce 's placed cm the operation of the Stand-or *;as Treat=ent System, discuss present andicr prc-osed:ech-Inical sDeci':catins and a4-nins:rative or ,hvs-.ca:controls 'rcv:iedd to ensure that these assu=;tions re-Main va-i-d.-here reliance 4s placed on other site-s;ecfic :con-siderations (e.g., refueling sequencing), prcvide 7resen:Cr ;rocosed technical specIficaticns, and tiscuss ad=nn:s-:ra::ve r. phvs ical controls provided to ensure the valid-4:y of such ccnsiderat;cns.-. Anal,'rses zer:n.ned to demonstrate wh 'Cri:erla "t :nrcuzh shculi conform to the guidelines c' NT7E, Q1/21.Atoendix A. .nus:ifv a exception taken tc -iese g-. ..el.nes,and srovice the s;.ecif'c information re'uestet d At:aznen.. -.3, or ., as a':ro;riate, for each analysis rerfc.-e=:.". 2 STEC!FIC RE' P.9 S FOR OVERH.EAD EN.LING SYSTD'.S OP-.-K: :N ?:N S:O~qAINlNG RKrQU -FOR P.EACTOP. C DE HY EL-.~OR 5?=L "T POOL COOLI;GRU'EG Section 5.1.5, -rovides guidelines concerming the designan! :.eration of load-handling svste=s in the vicinity of t:ui;nen: or co-7-nenzs re;uired for safe reactor and decay heat removal. inf c._a-:>~. 6rovied in respo.se to this section should be suffi-ien: to :emonst:a:t:'a: adecqate measures have been taken to ensure that in i:.ese areas, efther:h.e likelihood of a load drop which might prevent safe reactor shu:ooc orcontinued decay heat removal is extremely smal:, or that danaee tos cuipment from load drops will be li=ited in order not to resui.: rn toeoss of these safe:y-related functions. Cranes which =ust be evaluate- inSect:cn have betn previously identified in your res;onse *o 2. -., and:7eir :ca=s -in "ocur response to 2.1-3-c..:den:ifK anv cranes listed in 2.1-1, above, which vFu xaveevaluated as having sufficient design features to :.akt !-kelihood of a load drop ex:remely mall for all :oa;s :-e zarried and the basis for this evaluation (.e.. cc=:7.:E=:Tl1ance with %2E..G Ot12, Section or ;ar"+/-al cP-:,iance su ;lenen:ed by suitable alternative or additicna:design features). For each crane so evaluated, ;rovide :,e1cad-hand1'-n2-sV'ster. (i.e.. ticn S;ecilfled in Atta.:hzent 1.
2. For any cranes identified in 2.1-1 no: as single-failure-procf in 2.3-1, a ccmprehensive hazard e-aiua:ronshould be prcvided which includes the follcwing +/-n -::en.a. The ;resentaticn in a =a:rix for--a: of all heavvloads and pc:ential i.-act areas where da-agenigh: occur to safety-related equipoen:. Heav"loads identifica:i4n should include designationand weight or cross-reference to !nfor--aticn ;ro-v!ded in 2.1-3-c. Inpact areas should be i-denti-fied by construction zones and elevations or bysome other =ethcd such that the area canbe located on the plant general arrangementdra'ings. Figure 1 provides a typical matrix.b. For each interaction identified, indicate whichof the load and i--pact area co-bInativns can )ebecause of separation and redundanctof safery-related equip-ment, mechanical sC:,sand/or electrical interlocks, or other st:e-specific considera:ions. Elimina:ion on thebasis cf the aforementicned consideration shouldbe SuFlemen:ed 'y the following specifiL inf=zra-tion:(1) For load/target combinatiors elimina:edbecause of separation and redundancy cfsafety-related equipment, discuss thebasis for determining that load dropsillinot affect continued svs:em v-Qera-tion (i.e., the abilit:' of :he syse-:to perform its safety-related func:ion'.(2) `here mechanical stops or electricalinterlocks are to be provided, presentdetails showing the areas where cranetravel will be prohibited. Addirtonal-i. provide a discussion concerning theprocedures that are to be used forauthorizing the bypassing of interlocksor removable stops, for verifying thatinterlocks are functional prior to craneuse, and for verifying that interlocksare restored to o;erabilitv after opera-:+/-ons which require bypassing have beencompleted.(3) ;;here load/target cot 'inations are eli=-inated on the .%asis of other, si:e-s:ec-4f"c censidera:ions (e.g.. =aintenancesequencing), provide present and/cr ;ro-posed technical specifications and dis-cuss ad--inistrative procedures or phvsi-cal cons:rain:s izvoked to ensure thevalidity of such considerations.--i- c. For interactions not eliminated by the analysis of2.3-2-b. above, identif7 any handling systems forspecific loads which you have evaluated as havingsufficient 2esign features to =ake the likelihoodof a load drop ex:re=e2y small and the basis forthis evaluation (i.e., complete compliance witHNUREG 0612, SectiOn 3.l,, or partial cor=liancesupplemented by suitable alternative or addition-al design features). For each so evaluated, pro-vide the load-handling-system (i.e.. crane-load-combination) information specified in Attacl.nent 1.d. For interactions not eli=inated in 2.3-2-b or 2.3-2-c, above, demonstrate using appropriate analysisthat damage would not preclude operation of suffi-cient equipment to allow the system to perform itssafety function following a load drop (KUM7J 0612.Section 5.1, Criterion IV). For each analysis soconducted, the following information should beprovided:(1) An indication of whether or not, for thespecific load baing investigated, theoverhead crane-handling system is designedand constructed such that the hoistingsystem will retain its load in the eventof seismic accelerations equivalent tothose of a safe shutdown earthquake (SSE).(2) The basis for any exceptions taken to theanalytical guidelines of NU.REG 0612, Ap-pendix A.(3) The information requested in Attachment 4.~
Nc:Es7 TO FI7CUE iNote 1: Indicate ty st-bols :ýe sarezv-relared e~i-j=ent. The licenseeshould provide a list consistent with the clarifiza:tin przvilejin 1.2-3.Note 2: Fazarl Eii=ination Categoriesa. Crane travel for this area/load combination prohibitedbv electrical interlocks or mezhanical step&s.b. System redundancy and separation precludes loss ofcaaabil iry of syste= to perform its safetv-rela:edfunction following this load drop in this area.c. Si:e-specifi: considerations eli=ina:e the nee- to con-sider !oad/equip=ent co=bination.d. Likelihood of handling syste= failure f:r :his i>ad isextremelv s=a7i (:.e. section 5.1.6 tE3 J'2e. Ana2ysis demcnstrates tha: zrane failu.re an: !oa. dropwill no- :a..age safety-related eqi.ipment.
I KAMJ~ ITypical lo~cd/IujiadI Area MIUrfxClAPIR: (iUr~tl~rl TIM CROOKS 01 N.$J ANDI EtAUIrKYW? Ilteala)Ux'AT I ull10$'irATR nif NuILDIMUIS) CONISEPwhIVIM; TUIl, IIO'Air APIAMS FiAAIrLP: RYA1IIl~ OIIlIIM:. AUXILIARY OUILUJIM.(laMIvraf ARIA ST LUS.SrUCTBUM WIJNKS)FJIVAT 100SAFETY -RAlATEDMW IPHISTIALAI.D MIININATIUUCATNU)SEIL.EATIO(SAFETY- IFATF.DnpI nwjnHA1AND FL.INIEATI441(A? iiAM-, t I- I -I---,alm lwallu lloatme)auto ISes, Itu..,, Ls*ad fdeastiII-Catio abouheld locludsdesisemtime mad veight)%post Fuel CooknuI 10/14 (100 me..)I 4 I I ________-96 -I- -4 4---- ------ a ------------- ----. I
S:%3LE-FALUjRE -PR0OO qANJLING SYSTEMSI. Provide the name of the manufacturer and the design-rated load ,W. If:he =axi=L critical load N'CO), as defined in NrREG 0553, is nc: the sameas the DRL, provide this capacity.2. Provide a detailed evaluation of the overhead handling systen with respectto the features of design, fabrication, inspection, testing, and operationas delineated in NUREG 0554 and supplemented by the identified alternativesspecified in NUREG 0612, Appendix C. This evaluation zust include a point-by-point comparison for each section of NUREC 0554. If the alternativeszf N'REG 0612, Appendix C, are used for certain applications in lieu ofcomplying with the reco.-endation of NTREC 0554, this should be explicitlvstated. If an alternative to any of those contained in N'RE5 D554 or NUREG0612, Appendix C, is proposed, details must be provided on the proposedalternative to demonstrate its equivalency.l/3. respect to the seismic analysis employed to demonstrate :hat the over-head handling system can retain the load during a seismic event equal to asafe shutdown earthquake. provide a description of the method of analysis,:he assumptions used, and the mathematical model evaluated in the analvsis..he description of assumptions should include :he basis for selection of:troliley an. load position.A. Provide an evaluation of the lifting devices for each single-failure-proofhandling system with respect to the guidelines of XTREG 0612, Section 5.1.6.5. Provide an evaluation of the interfacing lift points with respect to theguidelines of N12ECEG 0612, Section 5.1.6.1/ if the crane in question nas previously been approved by tne staff as satisfyingVREG 0554, Reg. Guide 1.104, or Part 3 to 2T0-AS09-1, please reference theaate of t-e staff's safety evaluation report or approval letter in liew ;fproviding the information requested by item 2.
ý = -: : .. ...... ...--... .. .. ... _ _ -1 --, -.... T .... .... I ... , _ _-- ' Ii.... .......... ...... .. .. .....Ni ....l I I ' MA'z.YS:5 OF R:ZLoG:2AL PELEzEEThe f --ving nr;rI.a:ior. sh..d be ;rov"ied fcr an analvsis ccn_'du:ed toCEzmnstra:e cot7fianze with Cri:er4on 1 of N .REZ; 0612, Sec:ion 5.1....I.IAL CO....SiASSL.7TcN$a. ldentif-.* the time after shutdown, the number of fuelassemblies damaged. and the assumed curation of radio-lcgical release associated with eacn accident analvzed.b. NL2EG 0612, Table 2.1-2, prcvides the asaumptions usedto arrive at generic conclusions concerning radiolcgicaldose consezuences. To rely on the radlological doseanalysis of NUREC 0612, the licensee should ".'erifv Thatthese assunD:iors are zonservat4.,i 1:1:h regar2 :t thePlant/siTe evaluated. if the assume:ions are noc con-seetva4-e for the pe: ific 7lant, or if a =cre site-specific analysis is required, the licensee shou! 2identifv plant-s-ecific assumptions used in place cfthose tab4lazed.c. Identify and provide the basis (e.e., VSNRC RegulatoryGuide 1.25) for any assu=ptions employed in site-specificanalyses not identified in KUREG 0612, Table 2.1-2.d. Dose calculations based on the termination or mi:iga:ionof radtolouical releases should be supported bv inf:--ra-tion sufficient tc demonstrate both that the ti=e ýelavassuzed is conserva:ive and that the syste-_ p:cvided toaccomplish such termnna:ion or mitigation will :erformits safety function jpon demand (i.e.. tne system meetsthe criteria for an Engineered Safety Feature). Specificinfor-mation so proviced should include the follow:ng:(1) Details concerning the loca:ion of accidentsensors, parameters zonitcred and the valuescf these parameters at which a safety signalwill be initiated, sys:e= response t Ime(Including valve-operation time), and thetotal ti=e required to auto=atically shiftfro= nor--al operation to isolation or filtra-tion following an accident.(2) A description of the ins:rumenta:ion and con-trols associated with the Engineered Safer:Feature which includes Infcrmation sufficien:to dencnstrate :h;.z the re;jire=ents (Secticn 4)of 1EEE 279-1971, "Criteria for ProtectionSyste=s for Nuclear Power Generating Stations,"are satisfIed.7T..
(3) A description of any Engineered SafetyFeature filter system which includes infor-mation sufficient to demonstrate compliancewith the guidelines of USNRC RegulatoryGuide 1.52, "Design, Testing, and MaintenanceCriteria for Engineered Safety Feature Atmos-phere Cleanup System Air Filtration andAbsorption Units of Light-Water-CooledNuclear Power Plants."(4) A discussion of any initial conditions,e.g., manual valves lo:ked shut, containmentairlocks or equipment hatches shut) necessaryto ensure that releases will be terminated ormitigated upon Engineered Safety Featureactuation and the measures employed (i.e., Tech-nical Specification and administrative controls)to ensure that these initial conditions aresatisfied and that Engineered Safety Featuresystems are operable prior to the load lift.2. METHOD OF ANALYSISDiscuss the method of analysis used to demonstrate that post-accident dosewill be well within 10CFM00 limits. In presenting methodology used indetermining the radiological consequences, the following informaticn shouldbe provided.a. A description of the mathematical or physical modelemployed.b. An identification and sumary of any computer programused in this analysis.c. The consideration of uncertainties in calculationalmethods, equipment perfor=-ance, instrumentationresponse characteristics, or other indeterminateeffects taken into account in the evaluation of theresults.3. CONCLUSIONProvide an evaluation comparine the results of the analysis to Cri:ericn io,7 'REC 0612, Section 5.1. If the postulated heavv-load-dr:-p a:ccen:a.alyzed bounds other -cs:-lated heavy-load drops, a lisL cf these boundedheavy loads.should be provided.* , :71 + .*., Aw. M 5.7--W WUT.'7". -.... L ..7 *N O M41 -..UM.M
Attachment (3)CRITICALITY ANALYSISThe following information should be provided for analysis conducted to demon-strata compliance with Criterion II of NUMEG 0612, Section 5.11. INITIAL CON'DITIONS/ASSLWTIONSThe conclusions of NUR.G 0612, Section 2.2, are based on a particularmodel fuel assembly. If a licensee uses the results of Section 2.2rather than performing an independent neutronics analysis, the assump-tions should be verified to be compatible with plant-specific design.For any analysis conducted, the following assumptions should be providedas a minimum:a. Water/UO2 volume ratiob. The boron concentration for the refueling waterand spent-fuel poolc. The amount of neutron poison in the fueld. Fuel enrichmente. The reactivity insertion value due to crushing ofthe coref. T-he kIfc value allowed by technical specificationsfor t~e core during refueling2. .MTHOD OF ANALYSISProvide the method of analysis used to dezonstrate that accidentaldropping of& heavy load does not result in a configuration of the fuelsuch that keff is larger than 0.95. The discussion of the method ofanalysis should include the following infortation:a. Identification of the computer codes employedb. A discussion of allowances or compensation forcalculation and physical uncertainties3. CONCLUSIONProv.de an evaluation co=paring the results of the analysis to Criterion IIof LKUREG 0612, Section 5.1. If the postulated heavy-load-drop accident3-1MM-,
bounds other postulated heavy-load drops, a list of these bo'--.ded heavyloads should be ;rovided.
......... ......, .- ..... ..... -.. ... ........II .......... ... ....... .-i i -Attacnr-nent (4)ANALYSIS OF PLANT STRUCTURESThe following infor--ation should be provided for analyses conducted to demon-strate co=pliance with Criteria !I! and IV of N'*REC 0612, Section 5.1.I. !NITILAL CONDITIONS/ASSL 57TIONSDiscuss the assumptions used in the analysis, including:a. Weight of heavy loadb. Impact area of loadc. Drop heightd. Drop locatione. regarding credit taken in the analysis fcrthe action of i=pact limitersf. Thickness of-walls or floor slabs impactedg. Assutotions regarding drag forces caused by theenvironmenth. Load combinations consideredi. Material properties of steel and concrete2. ~~A.0' OF ANALYSIS?rcvide the method cf analysis used to demonstrate that sufficien: load-carrying capability exists within the wall(s) or floor identifyany co=puter codes enployed, and provide a description of their capa.ilities.If test data was employed, provide it and describe its applicabi2ity.3. CONCLUSION?rovide an evaluation comparing the results cf this analysis with CriteriaIII and iV of KnEC 0612, Section 5.1. '*here safe-shutdow- eq-Ipoent hasa ceiling or wall separating it fro= an overhead handling syste=, pr:':idean evalua:ica to demonstrate that postulated load drops do not ;tne:ratethe ceiling or cause secondary missiles that could prever: a s.: ;te= from perfor'.ing its safety function.TM OT
J_I of 8SHIELDED SHIPP-I.,5 CASKSFOR NK"LEAR P0PERCERT, PLA.NT$SI -Fuel (Npwi and Spent)CEtR. 4986545058C5593159386078b 0 662736375LiL-1, 2, 3, 1.ICC, 1. 2, 3Va.-denburghXTS Model 10092 7 Cl927CI348 (Series)?3_27l'.APY LICENS-ESGeneral Elec:ric Co.Westinghouse ElectricChez-Nuclear Systems,Inc.Nuclear Fuel ServicesCobustlon Engineer-Uabcock & Wilcox Co.Che-Nuclear Syste=s,Inc.Westinghouse Elec:ticCo.Nuclear Fuel Services,inc.Cenera1 Electic Co.)7" Industries, In:.General Electric Co.CROSS LOT INUS. AP3.70,000126,20048,000620070006940450067,05045,00050,000ayE?, DLCAPC, CPL, DLP, DPC,TPL, FPC, JC?, %%?P,yE?ac, PCEPECAPL:PC. F?CVt?APO, 3EC,14'L, TPC,nXA ?C.Pry ,, TVA,AP.-. CPC,".TC, '??.BCE, XYC, .?-.l.n'^ .E,6, 4 01 Su;er Tiger669r,F? ,SCE,DCJC?.NS?,ZL?,900190.1IF 300N- LO12Cr-160,)47,50023,000CPL."ArC,CpC.Cpc.E c,C.'E%.S?,C?!.,P7:.~f-' A.- --
7-772 of 6SHIELDED SiP;!NuG CASPS CERTIFICATEDF N.LEAR POWER PLANTSII -'asteCERT. mOw_'/'PRIYJ'.AY LIC-NSrE-GROSS LOT INLES. (APWFoX.)S %CO-'DY LICE--NSE'5026BC-48-220605861446244B3-16144C'he.=-Nuiclear Syste~s,Inc.Nuclear E--Sneering Co.Nuclear Lngineering Co.Chem-N-jclear Syste=s,Inc.71,00030,00042,000AC,CYA,FC,? SA?L.MIC,AC,C? C,I:- C,Any,A.-C,FPLI%---_2BEC,DC,JC?,CPC,N?,7EC,Dn?,CPL,F? C,CL., Cv---,DLC, -PL,NPP, VZ.,DL?, IEl,NS?, ?CE,VP?C?.L,K.S?,GC,NSP,CEC,* .4.9'v~ r.3 C?,?EC,624446,0006272 d oly Pan:her6568 LL-60-1506574 le 200Nuclear Engineeriog Cc.Ternessee Valley Auth.Hitt=an Nuclear andDevelopnent Corp.Che=-N\clear Sys:ezs,Inc.61-00AFL., CC, DL?, KC73,00047,0006601LL-50-10070,000A? L,DLP,AC,CEC,F7rL,vz?iDLC,E C,EEC.CC,1-z,.Y A:--F- ,66` 9V!2 SuerTigerNuclear Engineering Cc.Tenessee Valley A-:h.45,00051,0036712 S-33-:50C: .-tnt.... ,.~------
3 of 6SMIELDED SHIPPING CASKS CERTIFICATEDFOR NUCLEAR POWER PLANTSII -WasteCz-n-& .Xc:)FLPRIXARY .LIC-SEECT.OS S LOT INLBS. (APPROX.)6744Poly TigerNuclear Engineering Co.Nuclear Enginaering Co.6771 SN-I9074A?-1009079 M;-100 Ser. 290o0 \-60090S6 Y. 100 Set. IHittman .Nuclear andDevelopment Corp.RPittman Nuclear andDevelopment Corp.Pittman Nuclear andDevelopment Corp.Pit:man Nuclear andDevelopment Corp.Pitt.an Nuclear andDevelopment Corp.HiBtman Nuclear andDevelopment Corp,.Cheer-Nuclear Systems,lnc.35,00060,00028,00098,00042,00046,000SECONARY LICI:ýSEtAPL, BEC, CC, DL?,)IC, NP?, Sm., vr?APL, CPC, DLP, '%P,SY., VEtDLCAPL,DLY,MCE,AL,N??,EGE,ILE,N?P,NCE,NNE,CEC, Cw-,JCP, -.%A,P ZcCEC,JC? IPEC,XY.'A,DL.-'nA.YACVt. f:RG E,9CS9 Y-\-IOCS9092 EN-3009093 L-40036,50043,00043,00056,500BGE, C'.E, CiC, :.=:,JCP, n'A, P??, ?ECMYAHYA9094CNSI-14-195-HAPC.M*E,DC,JCP,INS?PCC'VE?AMC,DC,JCP,APL,CYA,F'?L,O?,A?L,VPE,PEG,CE-,T.C,PEG,CL,T?C,VPEPCPL,CPC.,C?C,FEC,TV A,C".-,Op: .9096CNSI-2.1-300Chem-Nuclear Systems,Inc.57,450*sc.( At~a'.&Zlf 1-4s
4 of ESHIELDED SH;PPINS CAS.S CERTIFICATEDFOR NUCLEAR POWER PLANTS!I -WasteGRDSS LOT It.LBS. (APPROX.)CERT. 'ODL9105R:-LWaste CR.I9105 AL-33-90PRAk Y.- LICENSEEChie-Nuclear Systems,Inc.Che--Nuclear Systems,Inc.Cbe=-Nuclear Syste-s,Inc.Chem-Nuclear Systems.Inc.Che-Nuclear Syste-s,Inc.58,40041.300SECOV'DA.RY L:Cý;SEi *APC,A?C,DPC,h?P,AC,DPC,Vt?APC,CYA,C?C,91i39113C1,6-80A7-10051,500CL, DC, FL,G?C, JCP,VuCL,D.?C,NST,C'- ,FPC,F?C,C?-,CTC.pzC,7000F? C,V!N-2_,9122 ia--.5061,0003!CSei a :-e.
1~Attach.-.en (5)5 of 6SI"IELD7ED SHIPPING CASKS CERTIFICATEDFCk NUCLEAvR POUER PLAIJSIII -?yprodejctsCTR-1 .PRLV.RY LICENSEEEGROSS LOT INLES. (APPROX.)SECON&DRY LICENSEEPLC5971GE-2005980 .--60010,00018,50030,00026,000P.;E, NS?62759081LL-26-4CNS-1600Che--Nuclear Systems,Inc.Chem-Nuclear Syste=m,Inc.APC, CPL. DPC, FPL,FPC, N??, VE?APC. BGE, CL, DPC,FM1., FPC, GFC, NSP,TVA, VE?Seeof
~JAttaChment (5)L4CE':tE AStEVIA7IONS 6 of 6APC Alabama Power CompanyAPL Arkansas Power and Light CompanyBEC Boston Edison CompanyBGE Baltimore Gas and Electric CompanyCEC Consolidated Edison CompanyCPC Consu=ers Power CompanyCL Carolina Power and Light CompanyC1W Co-onwealth Edison CompanyCYA Connecticut Yankee Atomic Power CompanyDLC Duquesne Light CotpanyDLP Dairyland Power CooperativeDPC Duke Pover CompanyFPC Florida Power CorporationFPL Florida Power and Light CompanyGPC Georgia Power Co=panyIEL Iowa Electric Light and Power CompanyL Indiana and Michigan Electric CompanyJCP Jersey Central Power and Light-Company?CC Metropolitan Edison CompanyMYA Maine Yankee Atomic Power Companyh-vT Niagara Mohawk Power CorporationIN Northeast Nuclear Energy CompanyNP Nebraska Public Power CorporationNSP Northern States Power CompanyOPP Omaha Public Power DistrictPEI Philadelphia Electric CompanyPEG Public Service Electric and Gas CompanyPCC Portland General Electric CompanyPY Power Authority of the State of New YorkRGC Rochester Gas and Electric CorporationSnJ Sacramento Munici;al Utilities CorporactonTEC Toledo Edison Cor.,anyTVA Tennessee Valley AuthorityV`P Vir;inra Eletric and Power Co=panyVyC Vercnt Yankee Nuclear Power Cor4rastionYAC Yankee Atotic Zlectirc Co=parny~ Powe~r Cor.-any-~ iscS.:- p ~ cce Ccrporaticn