NRC Generic Letter 1980-11

NRC Generic Letter 1980-113, Control of Heavy Loads.

ML071080219
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Sterling, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, 05000363, 05000000, 05000355, Zimmer, 05000471, Washington Public Power Supply System, Shoreham, Satsop, Trojan, Bailly, Allens Creek, Cherokee, Marble Hill, Hartsville, Phipps Bend, Yellow Creek
Issue date:	12/22/1980
From:	Eisenhut D Office of Nuclear Reactor Regulation
To:
References
GL-81-007, NUDOCS 8103190732
Download: ML071080219 (25)

UNITED STATES NUCLEAR REGULATORY

COMMISSION

bVASHItNGTON.

0. C. 20555 December 22, 1980 TO ALL LICENSEES

OF OPERATING

PLAtNTS AND , , -

FOR OPERATING

LICENSES A.D -" HOLDERS OF CONSTRUCTION, PERMITS*Gentlemen:

Subject: Control of Heavy Loads in January 1978, the ',RC published NUREG-04l entitled, Progran -or the Resolution of Generic Issues Related to Nuclear Power 'lants -Report to Concress." As part of this procra,-, the TasL' Action Plan fzr Unresolved Safety Issue Task 'N;c. 4-36, "Control of Heavy Loads Near Spent Fuel," was issued.ý..!e have completed our review of load handlinc operatiors at nuclear Dower plants. A report describing the results of this review has been issued as NUREG-0612, "Zontrol of Heavy Loads at %uclear Power plants -Resolution of TAP A-36." This -eport contains several recornendations to be implemented by all licensees and applicants to ensure the safe handling of heavy loads.The purpose of this letter is to request that you review your controls for the handling of heavy loads to detern"ine the extent to which the cuidelines of Enclosure

1 are cresently satisfied at your facility, and:0 identify the changes and ,o-ifications that would be -equired in order to fully satisfy these guidelines.

To expedite your compliance wit" this request, we have enclosed the following: ,UREG-0'12, "Control of Heavy Loads at '-uclear Power Plants'" Enc'esure 1).Staff Position -Interim Actions for Control of Heavy Loads (Enclosure

2).Request for Additional Infor-,ation on Control of Heavy Loads (Enclosure wVith the exception of licensees for Indian Point 2 and 3, Zion I and 2 and Three Mile Island l (7These Ywere previously sent a letter)~1O&7 3.~-*, A'

I WOMPt--d=

-WMWWMWWW1MWddý

-2 -December 22, 1980 You are requested to implement the interim actions described in Enclosure 2 as soon as possible but no later than 90 days from the date of this letter.In order to enable the NRC to determine whether operating licenses should be modified (10 CFR 50.54(f)), operating reactor licensees are requested to provide the following:

1. Submit a report documenting the results of your review and the required changes and modifications.

This report should include the information identified in Sections 2.1 through 2.4 of Enclosure

3, on how the guidelines of NUREG-0612 will be satisfied.

This report should be submitted in two parts according to the following schedule:-Submit the Section 2.1 information within six months from the date of this letter.-Submit the Sections 2.2, 2.3 and 2.4 information within nine months.2. Furnish confirmation within six months that imc'zmentation of those changes and modi'ications you find are -essary will commence 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 2.4 for the above report.Furnish justification within six months for any changes or modifications that would be required to fully satisfy the guidelines of Enclosure

1 which you believe are not necessary.

T-h :riteria in NUREG-0612 are also applicable to applicants for operating ii:e-ses.

Such applicants are expected to provide the information re:ues:ed by item 1 above and to meet the same schedule of implementation as i;icated in 2 above. Any item for which the implementation date is Drio- to the expected date of issuance of an operating license will be ccnsicered 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

3 This requcst for information was approved by GAO under a blanket clearance number R0072 which expires November 30, 1983. Comments on burden and duplication may be directed to the U.S. General Accounting Office, Regulatory Reports Review, Room 5106, 441 G Street, N.W., Washington, D.C. 20548.Sincerely, Darre .IG Eisenhut, Director Division o Licensing Enclosures:

1. NUREG-0612

2. Staff Position 3. Request for Additional Irformation cc: w/o Enclosure

(1)Service List.~ ..

ENCLOSURE

2 STAFF POSITION -INTERIM ACTIONS'FOR

CONTROL OF HEAVY LOADS (1) Safe load paths should be defined per the guidelines of Section 5.1.1(1) (See Enclosure

1);Procedures should be developed and implementea per the guidelines of Section 5.1.1(2) (See Enclosure I);(3) -rane operators should be trained, qualified and conduct themselves per the guidelines of Section 5.1.1(3) (See'Enclosure

1);Cranes should be inspected, tested, and maintained in accordance with the guidelines of Section 5.1.1(6) (See Enclosure

1), and (5) In addition to the above, special attention should' e given to procedures, equipment, and personnel for the handling of heavy loads over the core, such as vessel internals or vessel insoection tools. This special review should include the following for these loads: (1) review of procedures for installation of rigging or lifting devices and movement of the load to assure that sufficient detail 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 that coulch lead to failure of the component;

(3) appropriate repair and replacement of defective components;

and (4) verify that the crane operators have been properly trained and are familiar with specific procedures used in handling these loads, e.g., hand signals, conduct o' operations, and content of procedures.

REQUEST FCt ADDITIO NAL I Nr0CR'T ION ýN CO;JROL OF HEAVY L"ýCUS 1. INTRODU:TION

Verificatian by the licensee that the risk associated with :oad-.and..ng failur,.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 a basis for the staff's review of the licer.see's evaluation.

Additionally, they have been 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 assIst in 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, Section o 1.6).b. The cznsecuences of a failure can be so',-.. to be acceD: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 ;eneric analyses reported in !:*'-rG 0612, requiring only verficaton that these generic asst-ptions are valid.for a secific site or erT.-ay a site-s.ecific analy,-sis.

3. re;ufred for safe shutdown and continued he.a" heat removal are 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:iztec r.. .e;ula:nry Cui.de 1.:9. .osi-.-cr.

C.1 (except those or por: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 identifiable in 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 include licensed shipping casks -rDvided for the transportation of irradiated 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. 1 1(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) :s provided to indicate multiple-sITe use of these ship~ing zasks.-he results of the licensee's evaluation..

as reported in response -. this recues:. 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-M re-uct:in of the c:ten-ial hazard fro= load-handling-system failures" has been sa.'.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 whnere F=.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-e s ."-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 a F.-'& 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 1/2af d 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 your e-!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 clearly Identify 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 that ioad-handle=g cperations remaim within safe load pat.hs, in:lud--mg

rocedures, if any, for dev-a

+/-.=z f!.m these paths.c. A tabula:icm of heavy loads to be handled by each crane which includes, the load loa-, weight, its designated lifting device, and verifi-cation that the handling of such load is gcverned by a written procedure containng., as a =i=i==, the infor--ati=-

Iden: 4 fied in XELG 0612, 5ect.ion 5.1.1(2).d. Verification that devices identifled in -. -.1 3-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 supplemented by X-7E D612., Section 5.1.1(4) or 5.1.1(5).

are not met. describe any ;roposed alternatives and demon-strate t.heL- ,;u:vaiency in ter.ms of load-handling rellabil+/-

i1..e.

tha ANSI 330.2-1976, ChaTer 2-2. has been invoked ,-ith respect to crane insnec:ic-_ -es:i-g.and zaintenance.

'-:ere any exceptio= .s take- -c :tis standard, suffic:--t should be provtded tc demrnstrate 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:i co- 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_5 se: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

z Carr.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. o into the reactor -.-essel or s'ent fuel szcraze -...I. e, ifv a--.; crazes :isted -n 2_2-1. atove, "hZ----Du

'-.ave evaluated as havi.-_. sufic:ient design feat-.-res

o =&;,e :-.e likelihood 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:_- -.e te- 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::-.e include 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 c can Ie rTeM.E- or b:.pas sed and the adninistrative

7r---Cuts in+/-,'oked tC ensure proper authcr aat:cn cr. 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;G RU'EG Section 5.1.5, -rovides guidelines concerming the design an! :.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 or continued decay heat removal is extremely smal:, or that danaee to s cuipment from load drops will be li=ited in order not to resui.: rn toe oss of these safe:y-related functions.

Cranes which =ust be evaluate- in Sect: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 xave evaluated 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 :,e 1cad-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:ron should be prcvided which includes the follcwing

+/-n -::en.a. The ;resentaticn in a =a:rix for--a: of all heavv loads and pc:ential i.-act areas where da-age nigh: occur to safety-related equipoen:.

Heav" loads identifica:i4n should include designation and 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 by some other =ethcd such that the area can be located on the plant general arrangement dra'ings.

Figure 1 provides a typical matrix.b. For each interaction identified, indicate which of the load and i--pact area co-bInativns can )e because of separation and redundanct of safery-related equip-ment, mechanical sC:,s and/or electrical interlocks, or other st:e-specific considera:ions.

Elimina:ion on the basis cf the aforementicned consideration should be SuFlemen:ed

'y the following specifiL inf=zra-tion: (1) For load/target combinatiors elimina:ed because of separation and redundancy cf safety-related equipment, discuss the basis for determining that load drops illinot 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 electrical interlocks are to be provided, present details showing the areas where crane travel will be prohibited.

Addirtonal- i. provide a discussion concerning the procedures that are to be used for authorizing the bypassing of interlocks or removable stops, for verifying that interlocks are functional prior to crane use, and for verifying that interlocks are restored to o;erabilitv after opera-:+/-ons which require bypassing have been completed.

(3) ;;here load/target cot 'inations are eli=-inated on the .%asis of other, si:e-s:ec-

4f"c censidera:ions (e.g.. =aintenance sequencing), provide present and/cr ;ro-posed technical specifications and dis-cuss ad--inistrative procedures or phvsi-cal cons:rain:s izvoked to ensure the validity of such considerations.

--i- c. For interactions not eliminated by the analysis of 2.3-2-b. above, identif7 any handling systems for specific loads which you have evaluated as having sufficient

2esign features to =ake the likelihood of a load drop ex:re=e2y small and the basis for this evaluation (i.e., complete compliance witH NUREG 0612, SectiOn 3.l,, or partial cor=liance supplemented 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 analysis that damage would not preclude operation of suffi-cient equipment to allow the system to perform its safety function following a load drop (KUM7J 0612.Section 5.1, Criterion IV). For each analysis so conducted, the following information should be provided: (1) An indication of whether or not, for the specific load baing investigated, the overhead crane-handling system is designed and constructed such that the hoisting system will retain its load in the event of seismic accelerations equivalent to those of a safe shutdown earthquake (SSE).(2) The basis for any exceptions taken to the analytical guidelines of NU.REG 0612, Ap-pendix A.(3) The information requested in Attachment

4.~

Nc:Es7 TO FI7CUE i Note 1: Indicate ty st-bols :ýe sarezv-relared e~i-j=ent.

The licensee should provide a list consistent with the clarifiza:tin przvilej in 1.2-3.Note 2: Fazarl Eii=ination Categories a. Crane travel for this area/load combination prohibited bv electrical interlocks or mezhanical step&s.b. System redundancy and separation precludes loss of caaabil iry of syste= to perform its safetv-rela:ed function 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 is extremelv s=a7i (:.e. section 5.1.6 tE3 J'2 e. Ana2ysis demcnstrates tha: zrane failu.re an: !oa. drop will no- :a..age safety-related eqi.ipment.

I KAMJ~ I Typical lo~cd/IujiadI

Area MIUrfx ClAPIR: (iUr~tl~rl TIM CROOKS 01 N.$J ANDI EtAUIrKYW?

Ilteala)Ux'AT I ull 10$'irATR

nif NuILDIMUIS)

CONISEPwhIVIM;

TUIl, IIO'Air APIAMS FiAAIrLP:

RYA1IIl~ OIIlIIM:.

AUXILIARY

OUILUJIM.(laMIvraf ARIA ST LUS.SrUCTBUM

WIJNKS)FJIVAT 100 SAFETY -RAlATED MW IPHIST IALAI.D MIININATIUU

CATNU)S EIL.EATIO(SAFETY- IFATF.D npI nwjn HA1AND FL.INIEATI441 (A? iiAM-, t I- I -I---,alm lwallu lloatme)auto I Ses, I tu..,, Ls*ad fdeastiII-

Catio abouheld locluds desisemtime mad veight)%post Fuel Cook nuI 10/14 (100 me..)I 4 I I ________-9 6 -I- -4 4---- ------ a -------------

---

. I

S:%3LE-FALUjRE -PR0OO qANJLING SYSTEMS I. 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 same as the DRL, provide this capacity.2. Provide a detailed evaluation of the overhead handling systen with respect to the features of design, fabrication, inspection, testing, and operation as delineated in NUREG 0554 and supplemented by the identified alternatives specified in NUREG 0612, Appendix C. This evaluation zust include a point-by-point comparison for each section of NUREC 0554. If the alternatives zf N'REG 0612, Appendix C, are used for certain applications in lieu of complying with the reco.-endation of NTREC 0554, this should be explicitlv stated. If an alternative to any of those contained in N'RE5 D554 or NUREG 0612, Appendix C, is proposed, details must be provided on the proposed alternative 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 a safe 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-proof handling 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 the guidelines of N12ECEG 0612, Section 5.1.6.1/ if the crane in question nas previously been approved by tne staff as satisfying VREG 0554, Reg. Guide 1.104, or Part 3 to 2T0-AS09-1, please reference the aate of t-e staff's safety evaluation report or approval letter in liew ;f providing the information requested by item 2.

ý = -: : .. ...... ...--... .. .. ... _ _ -1 --, -.... T .... .... I ... , _ _-- ' Ii.... .......... ...... .. .. .....Ni ....l I I ' M A'z.YS:5 OF R:ZLoG:2AL

PELEzEE The f --ving nr;rI.a:ior.

sh..d be ;rov"ied fcr an analvsis ccn_'du:ed to CEzmnstra: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 fuel assemblies damaged. and the assumed curation of radio-lcgical release associated with eacn accident analvzed.b. NL2EG 0612, Table 2.1-2, prcvides the asaumptions used to arrive at generic conclusions concerning radiolcgical dose consezuences.

To rely on the radlological dose analysis of NUREC 0612, the licensee should ".'erifv That these assunD:iors are zonservat4.,i

1:1:h regar2 :t the Plant/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! 2 identifv plant-s-ecific assumptions used in place cf those tab4lazed.

c. Identify and provide the basis (e.e., VSNRC Regulatory Guide 1.25) for any assu=ptions employed in site-specific analyses not identified in KUREG 0612, Table 2.1-2.d. Dose calculations based on the termination or mi:iga:ion of radtolouical releases should be supported bv inf:--ra-tion sufficient tc demonstrate both that the ti=e ýelav assuzed is conserva:ive and that the syste-_ p:cvided to accomplish such termnna:ion or mitigation will :erform its safety function jpon demand (i.e.. tne system meets the criteria for an Engineered Safety Feature).

Specific infor-mation so proviced should include the follow:ng:

(1) Details concerning the loca:ion of accident sensors, parameters zonitcred and the values cf these parameters at which a safety signal will be initiated, sys:e= response t Ime (Including valve-operation time), and the total ti=e required to auto=atically shift fro= 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 Protection Syste=s for Nuclear Power Generating Stations," are satisfIed.

7T..

(3) A description of any Engineered Safety Feature filter system which includes infor-mation sufficient to demonstrate compliance with the guidelines of USNRC Regulatory Guide 1.52, "Design, Testing, and Maintenance Criteria for Engineered Safety Feature Atmos-phere Cleanup System Air Filtration and Absorption Units of Light-Water-Cooled Nuclear Power Plants." (4) A discussion of any initial conditions ,e.g., manual valves lo:ked shut, containment airlocks or equipment hatches shut) necessary to ensure that releases will be terminated or mitigated upon Engineered Safety Feature actuation and the measures employed (i.e., Tech-nical Specification and administrative controls)to ensure that these initial conditions are satisfied and that Engineered Safety Feature systems are operable prior to the load lift.2. METHOD OF ANALYSIS Discuss the method of analysis used to demonstrate that post-accident dose will be well within 10CFM00 limits. In presenting methodology used in determining the radiological consequences, the following informaticn should be provided.a. A description of the mathematical or physical model employed.b. An identification and sumary of any computer program used in this analysis.c. The consideration of uncertainties in calculational methods, equipment perfor=-ance, instrumentation response characteristics, or other indeterminate effects taken into account in the evaluation of the results.3. CONCLUSION

Provide an evaluation comparine the results of the analysis to Cri:ericn i o,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 bounded heavy loads.should be provided.* , :71 + .*., Aw. M 5.7--W WUT.'7". -.... L ..7 *N O M41 -..UM.M

Attachment

(3)CRITICALITY

ANALYSIS The following information should be provided for analysis conducted to demon-strata compliance with Criterion II of NUMEG 0612, Section 5.1 1. INITIAL CON'DITIONS/ASSLWTIONS

The conclusions of NUR.G 0612, Section 2.2, are based on a particular model fuel assembly.

If a licensee uses the results of Section 2.2 rather 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 provided as a minimum: a. Water/UO 2 volume ratio b. The boron concentration for the refueling water and spent-fuel pool c. The amount of neutron poison in the fuel d. Fuel enrichment e. The reactivity insertion value due to crushing of the core f. T-he kIfc value allowed by technical specifications for t~e core during refueling 2. .MTHOD OF ANALYSIS Provide the method of analysis used to dezonstrate that accidental dropping of& heavy load does not result in a configuration of the fuel such that keff is larger than 0.95. The discussion of the method of analysis should include the following infortation:

a. Identification of the computer codes employed b. A discussion of allowances or compensation for calculation and physical uncertainties

3. CONCLUSION

Prov.de an evaluation co=paring the results of the analysis to Criterion II of LKUREG 0612, Section 5.1. If the postulated heavy-load-drop accident 3-1 MM-,

bounds other postulated heavy-load drops, a list of these bo'--.ded heavy loads should be ;rovided.

.........

......, .- ..... ..... -.. ... ........II .......... ... ....... .-i i -Attacnr-nent

(4)ANALYSIS OF PLANT STRUCTURES

The 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

57TIONS Discuss the assumptions used in the analysis, including:

a. Weight of heavy load b. Impact area of load c. Drop height d. Drop location e.

regarding credit taken in the analysis fcr the action of i=pact limiters f. Thickness of-walls or floor slabs impacted g. Assutotions regarding drag forces caused by the environment h. Load combinations considered i. Material properties of steel and concrete 2. ~~A.0' OF ANALYSIS?rcvide the method cf analysis used to demonstrate that sufficien:

load-carrying capability exists within the wall(s) or floor identify any 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 Criteria III and iV of KnEC 0612, Section 5.1. '*here safe-shutdow- eq-Ipoent has a ceiling or wall separating it fro= an overhead handling syste=, pr:':ide an evalua:ica to demonstrate that postulated load drops do not ;tne:rate the ceiling or cause secondary missiles that could prever: a s.: ;te= from perfor'.ing its safety function.TM OT

J_I of 8 SHIELDED SHIPP-I.,5 CASKS FOR NK"LEAR P0PER CERT, PLA.NT$S I -Fuel (Npwi and Spent)CEtR. 4986 5450 58C5 5931 5938 6078 b 0 6 6273 6375 LiL-1, 2, 3, 1.ICC, 1. 2, 3 Va.-denburgh XTS Model 100 92 7 Cl 927CI 3 48 (Series)?3_2 7l'.APY LICENS-ES General Elec:ric Co.Westinghouse Electric Chez-Nuclear Systems, Inc.Nuclear Fuel Services Cobustlon Engineer-Uabcock & Wilcox Co.Che-Nuclear Syste=s, Inc.Westinghouse Elec:tic Co.Nuclear Fuel Services, inc.Cenera1 Electic Co.)7" Industries, In:.General Electric Co.CROSS LOT IN US. AP3.70,000 126,200 48,000 6200 7000 6940 4500 67,050 45,000 50,000 a yE?, DLC APC, CPL, DLP, DPC, TPL, FPC, JC?, %%?P, yE?ac, PCE PEC APL:PC. F?C Vt?APO, 3EC, 14'L, TPC, nXA ?C.Pry ,, TVA, AP.-. CPC,".TC, '??.BCE, XYC, .?-.l.n'^ .E, 6, 4 01 Su;er Tiger 669r, F? , SCE, DC JC?.NS?, ZL?, 9001 90.1 IF 300 N- LO12 Cr-160,)47,500 23,000 CPL."ArC, CpC.Cpc.E c, C.'E%.S?, C?!., P7:.~f-' A.- --

7-77 2 of 6 SHIELDED SiP;!NuG CASPS CERTIFICATED

F N.LEAR POWER PLANTS II -'aste CERT. mOw_'/'PRIYJ'.AY

LIC-NSrE-GROSS LOT IN LES. (APWFoX.)S %CO-'DY LICE--NSE'

5026 BC-48-220 6058 6144 6244 B3-1 6144 C'he.=-Nuiclear Syste~s, Inc.Nuclear E--Sneering Co.Nuclear Lngineering Co.Chem-N-jclear Syste=s, Inc.71,000 30,000 42,000 AC, CYA, FC,? S A?L.MIC, AC, C? C, I:- C, Any, A.-C, FPLI%---_2 BEC, 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, 6244 46,000 6272 d oly Pan:her 6568 LL-60-150 6574 le 200 Nuclear Engineeriog Cc.Ternessee Valley Auth.Hitt=an Nuclear and Developnent Corp.Che=-N\clear Sys:ezs, Inc.61-00 AFL., CC, DL?, KC 73,000 47,000 6601 LL-50-100 70,000 A? L, DLP, AC, CEC, F7rL, vz?i DLC, E C, EEC.CC, 1-z,.Y A:--F- , 66` 9 V!2 Suer Tiger Nuclear Engineering Cc.Tenessee Valley A-:h.45,000 51,003 6712 S-33-:50 C: .-tnt.... ,.~------

3 of 6 SMIELDED SHIPPING CASKS CERTIFICATED

FOR NUCLEAR POWER PLANTS II -Waste Cz-n-& .Xc:)FL PRIXARY .LIC-SEE CT.OS S LOT IN LBS. (APPROX.)6744 Poly Tiger Nuclear Engineering Co.Nuclear Enginaering Co.6771 SN-I 9074 A?-100 9079 M;-100 Ser. 2 90o0 \-600 90S6 Y. 100 Set. I Hittman .Nuclear and Development Corp.RPittman Nuclear and Development Corp.Pittman Nuclear and Development Corp.Pit:man Nuclear and Development Corp.Pitt.an Nuclear and Development Corp.HiBtman Nuclear and Development Corp,.Cheer-Nuclear Systems, lnc.35,000 60,000 28,000 98,000 42,000 46,000 SECONARY LICI:ýSEt APL, BEC, CC, DL?,)IC, NP?, Sm., vr?APL, CPC, DLP, '%P, SY., VEt DLC APL, DLY, MCE, AL, N??, EGE, ILE, N?P, NCE, NNE, CEC, Cw-, JCP, -.%A, P Zc CEC, JC? I PEC, XY.'A, DL.-'nA.YAC Vt. f: RG E, 9CS9 Y-\-IOCS 9092 EN-300 9093 L-400 36,500 43,000 43,000 56,500 BGE, C'.E, CiC, :.=:, JCP, n'A, P??, ?EC MYA HYA 9094 CNSI-14-195-H

APC.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, VPEP CPL, CPC., C?C, FEC, TV A, C".-, Op: .9096 CNSI-2.1-300

Chem-Nuclear Systems, Inc.57,450*sc.( At~a'.&Zlf

1-4s

4 of E SHIELDED SH;PPINS CAS.S CERTIFICATED

FOR NUCLEAR POWER PLANTS!I -Waste GRDSS LOT It.LBS. (APPROX.)CERT. 'ODL 9105 R:-LWaste CR.I 9105 AL-33-90 PRAk Y.- LICENSEE Chie-Nuclear Systems, Inc.Che--Nuclear Systems, Inc.Cbe=-Nuclear Syste-s, Inc.Chem-Nuclear Systems.Inc.Che-Nuclear Syste-s, Inc.58,400 41.300 SECOV'DA.RY

L:Cý;SEi *APC, A?C, DPC, h?P, AC, DPC, Vt?APC, CYA, C?C, 91i3 9113 C1,6-80A 7-100 51,500 CL, DC, FL, G?C, JCP, Vu CL, D.?C, NST, C'- , FPC, F?C, C?-, CTC.pzC, 7000 F? C, V!N-2_, 9122 ia--.50 61,000 3!C Sei a :-e.

1~Attach.-.en

(5)5 of 6 SI"IELD7ED

SHIPPING CASKS CERTIFICATED

FCk NUCLEAvR POUER PLAIJS III -?yprodejcts CTR-1 .PRLV.RY LICENSEEE GROSS LOT IN LES. (APPROX.)SECON&DRY

LICENSEE PLC 5971 GE-200 5980 .--600 10,000 18,500 30,000 26,000 P.;E, NS?6275 9081 LL-26-4 CNS-1600 Che--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?See of

~J AttaChment

(5)L4CE':tE AStEVIA7IONS

6 of 6 APC Alabama Power Company APL Arkansas Power and Light Company BEC Boston Edison Company BGE Baltimore Gas and Electric Company CEC Consolidated Edison Company CPC Consu=ers Power Company CL Carolina Power and Light Company C1W Co-onwealth Edison Company CYA Connecticut Yankee Atomic Power Company DLC Duquesne Light Cotpany DLP Dairyland Power Cooperative DPC Duke Pover Company FPC Florida Power Corporation FPL Florida Power and Light Company GPC Georgia Power Co=pany IEL Iowa Electric Light and Power Company L Indiana and Michigan Electric Company JCP Jersey Central Power and Light-Company

?CC Metropolitan Edison Company MYA Maine Yankee Atomic Power Company h-vT Niagara Mohawk Power Corporation IN Northeast Nuclear Energy Company NP Nebraska Public Power Corporation NSP Northern States Power Company OPP Omaha Public Power District PEI Philadelphia Electric Company PEG Public Service Electric and Gas Company PCC Portland General Electric Company PY Power Authority of the State of New York RGC Rochester Gas and Electric Corporation SnJ Sacramento Munici;al Utilities Corporacton TEC Toledo Edison Cor.,any TVA Tennessee Valley Authority V`P Vir;inra Eletric and Power Co=pany VyC Vercnt Yankee Nuclear Power Cor4rastion YAC Yankee Atotic Zlectirc Co=parny~ Powe~r Cor.-any-~ iscS.:- p ~ cce Ccrporaticn