ML20086P178
| ML20086P178 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/04/1984 |
| From: | George H, Hladik M GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20086P173 | List: |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR NUDOCS 8402270031 | |
| Download: ML20086P178 (94) | |
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EVALUATION OF HEAVY LOAD HANDLING OPERATIONS AT THREE MILE ISLAND - UNIT 1 RESPONSES TO NRC REQUESTS IN TER TRANSMITTED 10/21/81 P REP ARE D BY:
December, 1983 Mattnew Hladik Mechanical Components H.
George Project Manager Tera Corporation APPROVED BY:
h Y=-;,
B.
D.
Elam Date: /[4/8/
Manager Mechanical Components
. h-D.
K.
Crone ger Date: [. -
Director, Engineering & Design f
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- f ENCLOSURE 1 RE S PO NSES TO NRC REQU ES TS IN TER TRANSMI TTED 10/21/81 NRC Req ues.: Guideline 1 - Safe Loa d Pa th s "Sa f e -l oad pa th s should be d e fine d for the m ovemen t of heavy loads t o minimiz e the po ten tial f or heavy load s,
i f dropped, to impac t i r radia ted f uel in the reactor vessel and in th e spen t f uel pool, or t o impac t safe sh utdown eq uipmen t.
Th e pa th should follow, to the exten t pra c ti ca l, s truc tural flo or member s, b e am s,
e tc., ruch th a t i f the load is dropped, the s truc ture is more likely t o wi ths tand th e impact.
Th e s e load p a th s should be de fined in procedures, shown on eq uipm e n t layout drawings, and clearly marked on the floor in th e area where the load i s to be handled.
Devia ti ons fr om de fine d loa d pa th s should req uire wri tten alterna tive procedures approved by the plant sa fe t y revi e w c ommi t te e. "
RE S PO N S E :
Th e r e are two areas o f concern in thi s case: (a) the Reactor Building (wi th a 18 5 ton polar crane), and (b) a 110 ton overhead the Fu el Ha ndling Building (wi th cran e).
Reactor Building Load handling procedure s and eq uipment layout drawings have bee n amende d t o iden ti f y sa f e loa d pa th s.
Th e s e procedures contain a plant layout drawing whi ch is marked to clearly de fine th e sa f e loa d handling area for th e desi gna ted l oa d.
0012N.,-
Devi a ti on s Any devia tions from these procedures and load pa ths require a revision to procedures or a Temp ora ry Change No ti ce, e i th e r t o whi ch - m us t b e appr oved by the Plant Revie w Group and Opera tions and Main tenance Director,
( TMI-1 Pro cedure 10 01 A).
TMI-1 ha s endeav ore d to sa ti sf y th e intent of th e NkC guideline t o provide speci fic pa thways for the m ovem en t o f heavy loads th r o u gh the us e of procedures, a nd s ui table drawings.
Load pa ths have been chosen s o t o minimiz e potential sa fe ty hazard s due to heavy as load impact.
S tructural analyses as described in den ons tra te the accep tabili ty o f d esigna ted load pa th s.
physical markings Exceptions Since th e r e are different safety concerns f or ea ch of the heavy loads tha t m us t be handled by the Reactor Building Crane and th e r e a re a large varie ty o f heavy loads tha t m us t be handled, defining safe load pa ths in th e manner de s cribe d i n NUREG-0 612, Section 5.1.1(1), i s nei ther req uire d n or pruden t for every si tua tion.
To d o s o, such as t ryi n g to pu t markings on th e floor f o r e a ch l oa d, would cause unnecessary confusion.
l l
Th e si gnalman wi th the j ob s upervisor will inspect the desi gna te d loa d pa th pri or to a loa d m ovemen t in order t o a ssure th a t there are n o obs truc tions that could af fec t th e abili t y o f th e operator to follow th e d esi gna te d pa th.
0012N _
Fuel Ha n dli n g Building Adminis tra tive an d Physica l Con trol s The handling o f heavy load s in the fuel handling building is s tri c tly c ontr olled by a dminis trc tive guideline s an d limi ta tion s.
These include section 3.11 o f th e TM I -l Te chnical Spe ci fi ca ti ons and approved plant proce d ure s.
Along wi th th e s e, a ke y operated travel in terlock system which physically limi t s fuel handling building crane travel i s em ploye d when loads in e xcess o f 15 tons are to be handled.
Ce rtai n cha nge s to th e s e procedure s and th e us e of the travel in terlock sys tem are proposed a s described in.
Load Classes
[ A procedural change reques t has been s ubmi t ted to req uire a n iden ti fie d loa d pa th and evalua tion for loads in e xcess o f 15 tons.]
Th e only area of the spen t f uel pool s truc ture tha t may be exposed to a spent f uel cask accidentally dropped from a h e i gh t grea ter -tha n one foot is the shipping cask area whi ch
- h as been designed t o wi th s tand the impact from a dropped spen t f uel cask. (FSAR 9.7.1)
Loads weighing less than 3000 p ounds a re not subject to th e re s tri c ti on s of T. S. 3.11.6.
Thi s wa s found to be a cceptable because the consequences of dropping load s in thi s wei gh t range are c om pa ra ble to th os e produced by the f uel handling a ccident evalua ted in th e FSAR section 14.2.2 and f ound accep table.
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Procedural Req ui rem e n t s I tL i s th e re sponsibility of each and every individual involved in a heavy load lif t to understand and follow proced ure s.
Any devia ti on s from th e s e procedure s and l oa d p a th s re q uire a revision to procedures or a Tem p ora ry Change No tice, ei ther o f whi ch mus t b e
. ap proved by the Plant Review Group and Opera tions and Maintenance Director. ( TMI-l Procedure 10 01 A)
NRC Req ues t: Load Handling Procedures, (Guideline 2, NUREG-0612, Section 5.1.l(2)
" P r,o c e d u re s should be developed to cover load handling op era ti on s f or heavy loads tha t are or could be handled over or in proximity to irradia ted fuel or
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safe shutdown eq uipmen t.
At a minim um, procedures should cover handling of th o s e l oa ds ' 11 s ted i n Ta ble 3-1 o f NUREG-0 612.
The s e proced ure s should include:
i den ti fica tion o f req utre d eq uipmen t; inspections and accep tance cri teri a req uire d be f ore m ovemen t o f loa d; th e steps and proper seq uence to be f ollowed in handling th e loa d; de fining th e safe pa th; an d o th e r sp e cial pre cauti ons."
RE S PO N S E :
Li f ting Procedures Th e f ollowing li f ting procedures are used to handle heavy load s in th e - TMI-l Reactor Buildi n g:
1504-4 Rev.
8-In s talla ti on of Transfer Ca na l Sea l Pla te and Flood Line Cove r Pla te 0012N !
=.-_-
=
r.
1504-7J Rey, 4 -
Clocu're Hoa d Rem oval
~
e 15 04-8' Rev.
4-Re ac tor Upper Plenum Rem oval 1504-15 Rav
.5 -
Rem ova l of Core Support As sembly 1506-1 Rev.
3-Reactor Upper Plenum In s talla ti on
- 1506-2 Rev.
2-Closure Hea d Ins talla tion 1506-8 Rev.
3-S t orage of Transfer Canal Seal Pla te and Flood Line Cover Plate 1506-13 Rev. 2 -
Reins talla tion o f Core Support Assembly 1509 Rev. 0 -
Reactor Vessel Missile Shield Handling 1508-1 Rev.
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Incore Ins trum en ta ti on The se procedure s have been revised or will be revised to sa ti sf y ' th e req ui rem e n t s o f 5.1. l( 2) o f NUREG-0 612 I
i ncluding:
i den ti fi ca ti on o f proper handling i
. eq uipm en t; define d saf e load handling area s; pre ca u ti ons; p re req ui si te s; training and q uali fica tion req ui rem en t s for cr an e op era t or s ; ve ri fi ca ti o n th a t,
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where applicable, req uire d de tailed inspec tions have
-been p erf onae d; required crane 'in spec tion by opera tor
. pri or to handling; s upervi sion o f work involving a
heavy load li f t by a designa te d j ob supervis or; and the.s teps and proper seq uence to be f ollowed in handling the load.
0 012N.,
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Fue l - Handli n g Building Loa d Handling Handling o f ' he avy loads in the f uel handling building
- L s - n o t covered by a procedure f or ea ch load,
i.e.,
sp ent f uel shipping cask.
Instead, the se heavy loads are con trolle d by the admini s tra tive te chni ca l sp _
- ica tions a nd re f ueling procedure, 1507-2 Fuel Handli ng Building Crane Op era tion.
Thi s Procedure re q uire s - th a t' a " safe l oa d pa th" and evalua tion be made f or all load s in exces s of 15 tons pri or to handling the load.
Th e f ollowing i s a point by point evalua tion o f the heavy loads tha t could be or are handled by the TM I -1 f uel handling' building crane.
Spent Fue l Ship pi n g Ca sk s Spent fuel shipping cask s are no t p e rm i t te d i n the Unit 1 f uel handling building a t the present time due to NRC-prohibi tion. ( TS 3.11.7)
Thi s prohibi tion will
.b e re tained un til GPU N s ubmi ts a t e ch ni ca l spe ci fica tion amendmen t ap prove d by th e NRC, concluding th a t spent fuel cask handling opera tions a t TMI ca n b e p e rf orm e d sa f ely.
At tha t tim e th e safe l oad pa th for the spent fuel shipping c sks will be i
provide d by the ke y op era te d travel interlock sys tem
[
a nd ap prove d procedure s.
Pool Divider Ga t e Movement of the pool divider ga te, wei ghi ng j
ap proxima tely 5000 pound s, be twee n th e spen t fuel p ools. would req uire a ch ang e to the t e ch ni c al speci fica tion s and NRC ap proval.
Te chni ca l
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Specification 3.11.8 ha d previously waived the 3000 p ound t o 15 t on handling cri teria ( TS 3.11.6) prior to th e Cycle 5 re fueling outage only, provioe d certain cri teri a were m e t.
Ne w Fue l Shipping Con tainer s Handling o f new fuel shipping containers is con trolle d by Procedure 1507-2, Se etion 3 of th e TM I -l tech specs and Re fueling Procedure 1303-1, " Receipt, In spe c ti on,
Fi t-Up and S torage o f New Fuel and Con trol Componen ts".
~ Faile d Fue l Containers Th e wei gh t of a failed f uel con tainer wi th a f ue l i
a ssembly and handling mas t has been de termined to wei gh le s s - tha n 300 0 p ound s and, th e re f ore, is not considered a heavy load a t IMI.
It will be handled in accordance wi th te chni ca l speci fica tion and re f ueli ng procedure 1507-2.
Fue l Tran s f e r Ca r ri a g e Th e fuel transfer carriage and a ss ocia ted eq uipmen t at TMI-l a re ins talle d eq uipm en t, thes e would generally be re pair e d in place.
Sh ould maintenance or design m odi fica tion warran t li f ting or rem ov a l, special procedure s would be wri tten and approved governing m ovemen t nea r f ue l.
Ra di oa c tive Wa s te Shi pping Ca sk Handling a ra di oa c tive waste shipping cask in the fuel handling building i s no t a concern und e r NUREG-0 612 because the re sin cask i tself i s not lifted over safe shutdown eq uipmen t or irradia te d f ue l.
O th e r loads I
0 012N L
as socia te d wi th th e s e casks, su ch a s th e cask lid, are h andled in a ccordance wi th procedure 15 07-2 whi ch sa ti sfie s NURE G 0 612 concern s.
Pl an t 'Eq uipmen t If plant eq uipmen t m us t be handled and b rough t into th e plant th r o u gh th e f ue l handling building tr u ck bay, these will be handled per S e c ti on 3.11 of the TMI -l te chni ca l speci fica ti on s and Re fueling Procedure 1507-2.
Plant e q ui pm e n t would generally be brough t directly into th e buildi ng s erved.
No plant eq ui pm e n t was i den ti fied req uiring handling in th e TMI-l Fu el Han dli ng Buildi ng.
Ir radia ted Specimen Shipping Cask All ir radia ted sp ecimens have been rem oved f rom TM I -l and sen t of f site; th ere f ore, an irradia te d specime n shipping cask i s no t e xpec ted to be handled for TMI Uni t 1.
NRC Req ue s t:
Crane Opera tor Training (Guideline 3, i
NU REG-0 612, Article 5.1.l(3)
(
" Crane operators should be trained, q uali fied, and conduct them selve s i n acc ordanc e wi th Ch a p t e r 2-3 of ANSI B3 0. 2-19 7 6,
" Ove rhe a d a n d Ga n try Cr ane s. "
RE S PO N S E :
Maintenance Pr o cedure MP-14 06, " Crane Operator Quali fi ca ti on s" represents th e d o c um e n ta ti o n of th e crane operator training program f or THI-1.
Th e practice s de s cribe d in thi s procedure com ply wi th ANSI 0 012N - l
i B3 0.2-19 7 6, Chap t er 2-3.
Th e training of crane operators is c onduc ted by b oth the maintenance de pa rtm e n t and th e traini ng s ta f f.
Th e training s taf f is re sp onsible f or i ns truc tion d uring the c l a s s r o om p ortion o f the training and th e practical f ac t or s training is the re sp onsibili ty o f the approved Ins truc t or/ Quali fi er s from th e Main tenance De p a r tm e n t.
At TM I -1, the physical and training records for the crane opera tor are main taine d in th e following manner :
Th e Sa f e ty and He al th Department p e r f o rm s a require d annual physical for th e crane operators.
Th e crane opera tor physical i s given i n conj unction wi th the annua l re spira tor physical.
Th e re sults and re cord s of the crane operator physicals are maintained by th e Sa fe ty and Heal th Department.
In the event a crane operator fails th e physical, the ap propria te Main tenance Department In s tr uc tor / Quali fier is notified by Sa fe ty an d Heal th.
Th e Main tenance De pa rtmen t In s truc tor / Quali fier s
-are responsible f or giving th e crane operators th e p ra c ti cal factors portion of the training.
Th e shee t documen ti ng sa ti s fa c t ory or un sa ti s f a c tory p erf orm ance d uri ng the practical factors training i s th e Crane Operator Quali fi ca ti on Re cord.
Th i s record is c om ple te d and maintaine d by th e Ins truct or/Quali fier wi thin the Main tenance De partmen t. Th e In s truc tor /
Quali fi er als o prepare s and main tains a crane opera tor q ualifica tion ma tri x th a t lis ts all the crane operators, the cranes the y are q ualifie d on a nd the da te they were q uali fied.
A copy of thi s ma tri x i s sen t to th e Traini n g De p a r tm e n t for th e i r file s.
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Th e Training De pa rtm en t main tain s cla s s room training d ocum en ta ti on on th e crane operators along wi th the ma tri x supplie d by main tenance.
A copy of the q ualifica tion ma tri x will also be maintained a t th e Shi f t Maintenance Of fice.
MP-14 06 i s pre sen tly being revi sed to further sa ti sf y ANSI B3 0. 2-19 7 6 by adding th e f oll owi ng i tem s t o th e procedure:
An annual re training pro gram i s being develope d t o ensure th a t all crane operators remain q uali fie d.
However, in addi tion to the annual re training, the Manager of Plant Maintenance or his de si gne e shall req ui re th a t an operator previ ously designa ted a s q uali fied, who no longer possesses th e req ui si te proficiency b e subj ec t to re training a t any time.
MP-1406 will als o include in th e "Pequirements" s ec ti on, a f onn al. s ta tem en t tha t crane operators are ins truc te d on proper operator condec t per o
ANSI B3 0. 2-19 7 6.
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1 ITIC Request: (from draft Franklin TER)
A conclusion with respect to this guideline must be deferred until completion of the evaluation currently in progress.
Since the Licensee has indicated in preliminary review that EOCl specifications were met at the time of crane design, the Licensee should address design requirements invoked by CMAA-70 which are more restrictive than those invoked in the EOCi standard (e.g.,
allowable compressive stress in structural members).
Response
The requested verification is provided below for the Fuel Honoling Building Crane and the Reactor Building Crane.
Fuel Hcndlino Buildino Crane The Fuel Handling Building Crane was built prior to the issuance of ANSI B30.2-1976 and CMAA 70-197S. This crone was designed and fabricated by Whiting Crane Corporation in accordance with EOCl-61, " Specifications for Electric Overhead Traveling Cranes-1961," and additional criteria contained in Gilbert Associates Specification No. TMI-MG, Issue 3, July 29,1969. These specifica-tions oddressed certain, but not all, of the criteria in ANSI B30.2-1976 and CMAA 70-1975. To oddress the 10 points identified in the Franklin Ftesearch Institute's l
TER where CMAA-70 and ANSI B30.2 are more restrictive than EOCl-61, a design eva!uation of the Fuel Handling Building Crane was performed. The following summarizes our findings for these 10 points.
(l)
Torsional Forces - CMAA 70 specifies that twisting moments be determined based on the horizontal distance between the center of gravity and the shear center of the girder section. EOCl-61 requires twisting moments to be based on the distance between the load center of gravity and the beam center of gravity. Since the Fuel Handling Building Crane girders are symmetrical box sections, l
11
4 these two requiremen+s are the same. Since the trolley rails are located over the centerline of the girders, there are no torsional forces on the girders due to vertical loads. Thus Fuel Handling Building Crane satisfies CMAA %..
70 criteria relative to torsional forces.
(2)
Lonaltudinal Stiffeners - CMAA 70 specifies a minimum moment of inertia for longitudinal stiffeners, maximum width to thickness ratio, and stiffener location along the web plate. EOCl does not provide similar guidance. For the Fuel Handling Building Crane, application of the CMAA criteria requires that the moment of inertia be greater than lo = 2.20-in.4, the width to thickness ratio should be less than 12, and the stiffener shou.id be located 0.4 of the distance from the compression plate to the web neutral oxis. The actual moment of inertia is 2.23-in.4, the stiffener width to thickness ratio is 2.0, and the stiffener centerline is located 0.39 of the distance from the compression plate to the web neutral axis. Thus the CMAA criteria relative to longitudinal stiffeners cre satisfied for this crone.
CMAA 70 specifies that 1/h (i = girder spon; h = web height) should be less than 25; EOCl-61 has no limit on 1/h.
For tne Fuel Handling Building Crane,1/h = 568 in./70 in.
= 8.1. Therefore, CMAA 70 is satisfied.
In addition, CMAA 70 specifies that h/t be less than C(K+1) 17.6 and less than M, where:
3 fc web thickness = 5/16 in.
t
=
162 (the Fuel Handling Building Crane has one C
=
longitudinal stiffener) f /fc = 1.0 K
t
=
12
i ft
=
max. tensile stress = 16.0 ksi fe max. compressive stress = 16.0 ksi
=
M 376
=
Therefore according to CMAA 70, h/t should be less than 339.8 and less than 376. h/t = 70/(5/16) = 224. Therefore, CMAA 70 is satisfied.
(3)
Basic Allowable $iresses - EOCI-61 is more conservative than CMAA 70 for allowable tension, compression, and shear stresses, if b/c is less than 38 (b is distance between web plates and e is the thickness of the cover plate). For
-the Fuel Handling Building Crane, b/c is 11.5 in./.75 in. =
15.3. Therefore, CMAA 70 is satisfied.
(4)
Fatiave Failure and Cyclic Loading - CMAA 70 specifies that fatigue failure be considered in the crane design, and also specifies an allowable stress range for crane struc-tural members that are subject to cyclic loading of greater than 20,000 over the life of the crcne.
The number of cycles for any of the crane members will be less than 2,000 over the life of the Fuel Handling Building Crane. Based on this, failure due to cyclic fatigue should not be of concern for this crane, ar.d the CMAA 70 criteria for cyclic loading are satisfied.
(5)
Holstino Rope - CMAA 70 specifies a 5:1 hoisting rope j
safety factor for the rated load plus bottom block divided l
by the number of parts of rope. For the Fuel Handling L
Building Crane, the resulting safety factor for the main hoist is:
l l
l bottom block = 5,000 lbs.
rated load = 220,000 lbs.
parts of rope = 12 (I-l/8" each) 4 DC-83-25 4
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rope published breaking strength = 100,000 lbs.
resulting safety factor = 100,000/((220,000+5,000)/12)=5.33:1 For the aux. hoist:
,(
block = 2000 lbs.
rated load = 30,000 lbs.
parts of rope = 1" dio.,8 Ports rope published breaking strength = (g.t.) 20,000 lbs.
resulting safety factor = 20,000/(32,000/8) = 5.0:!
Therefore the ropes satisfy the criteria in CMAA 70.
(6)
Holst Drum Loads - CMAA 70 specifies that drum design should consider combined crushing and bending loads; however, EOCl 61 is not as specific. The Whiting standard design practice for this crone was to consider combined crushing and bending foods.
Therefore CMAA 70 is satisfied.
J (7)
Holst Drum Groove - CMAA 70 specifies minimum drum groove depth and drum groove pitch; EOCl 61 does not provide such specific guidance. - For the Fuel Handling Building Crane, Whiting Crone criteria required the mini-mum groove depth to be 3/8 x rope diameter and groove pitch to be rope diometer plus I/8 inch. The drum pitch diameter was required to be greater than 24 x the rope diarneter. These standard practices were later incorpo-rated into CMAA 70.
Thus the CMAA criteria are satisfied.
(8)
Holst Holdina Brakes - CMAA 70 and ANSI B30.2 require that holding brakes have minimum torque ratings (relative to motor torque) of 125% if used with control broking other than mechanical; and 100% if used with mechanical control broking.
14
l For the Fuel Handling Building Crone, eddy current con-trol braking is used. The main hoist holding broke is 300%
of motor torque, and the oux, hoist holding broke is 170%
of motor torque. Thus the CMAA criteria are satisfied.
i.
(9)
Static Controls - CMAA 70 includes various criteria for crane static controis; EOCl only addresses crone magnetic controls..Since the Fuel Handiing Building Crane uses a D.C. magnetic control system, the CMAA 70 criteria on static control are not applicable.
(10) Restort Protection - CMAA 70 establishes criterio for restart protection for crones not provided with spring-return controllers or momentary contact pushbuttons; this is not addressed in EOCl 61. The redundant controller for the Fuel Handling Building Crane uses switches that are spring return type. Crone controls in the cab are G.E.
l switches which are not spring returned. Although the control schematic does provide undervoltoge protection, devices are not provided which prevent motors from being restarted until the controlier handle is brought to the "OFF" position.
These switches will be modified to include a spring return to the "OFF" position.
Reactor Buildino Crane j
The Reactor Building Crane was built prior to the issuance of ANSI l
830.2-1976 and CMAA 70-1975.
This crone was designed and l-fabricated by Whiting Crane Corporation in accordance with EOCI-61,
" Specifications for Electric Overhead Traveling Cranes-1961," and additional criterio contained in Gilbert Associates Specification No.
l~
TMI-MG, Issue 6, March 3,1970.
These specifications addressed certain, but not all, of the criteria in ANSI B30.2-1976 and CMAA 70-1975. To address the 10 points identified in the Franklin Research j
institute's TER where CMAA-70 and ANSI B30.2 are more restrictive DC-83-25
~
15
=
than' EOCl-61, a design evaluation of the Reactor Building Crand was performed.
The following summarizes our findings for these 10 points.
(1)
"orsional Forces - CMAA 70 specifies that twisting moments be determined bcsed on the horizontal distance between the center of gravity and the shear center of the girder section. EOCl-61 requires twisting moments to be based on the distance between the load center of gravity and the beam center of gravity.
Since the Reactor Building Crone girders are symmetrical box sections, these two requirements are the same. Since the teolley rails are located over the centerline of the girders, there are no torsional forces on the girders due to vertical loads. Thus Reactor Building Crane satisfies CMAA 70 criteria relative to torsional forces.
(2)
Loncitudinal Stiffeners - CMAA 70 specifies a minimum moraent of inertia for longitudinal stiffeners, maximum width to thickness ratio, and stiffener location along the web plate. EOCl does not provide similar guidance. For the Reactor Building Crane, application of the CMAA criteria requires that the moment of inertia be greater than lo = 6.29-in.4, the width to thickness ratio should be less than 12, and the stiffener should be located 0.4 of the distance from the compression plate to the web neutral axis.
The actual moment of inertia is 9.25-in.4, the stiffener width to thickness ratio is 2.0, and the stiffener centerline is located 0.4l of the distance from the compression plate to the web neutral axis.
Thus the CMAA criteria relative to longitudinal stiffeners are satisfied for this crane.
CMAA 70 specifies that 1/h (l = girder span; h = web height) should be less than 25; EOCl-61 has no limit on 1/h.
16
. ~ -. - -. - -. - -. -.
For the Reactor Building Crane,1/h u = 14.4 Therefore, l
CMAA 70 is satisfied.
In addition, CMAA 70 specifies that h/t be less than i
CO<+1)
M and less than M, where:
T fc web thickness = 5/16 in.
t
=
162 (the Reactor Building Crone has one C
=
longitudinal stiffener) f /fc = 1.0 K
=
t max. tensile stress = 16.0 ksi ft
=
max. compressive stress 216.0 ksi fe
=
376 M
=
Therefore according to CMAA 70, h/t should be less than 104/(5/16) 333.
339.8 and less than 376.
h/t
=
=
Therefore, CMAA 70 is satisfled.
(3)
Basic _ Allowable Stresses - EOCl-61 is more conservative than CMAA 70 for allowable tension, compression, and shear stresses, if b/c is less than 38 (b is distance between web plates and c is the thickness of the cover plate). For the Reactor Building Crane, b/c is 27.5 in./2.0 in. = 15.3.
Therefore, CMAA 70 is satisfied.
l (4)
Fotiaue Failure and Cyclic Loading - CMAA 70 specifies that fatigue, failure be considered in the crane design, and also specifies an allowable stress range for crane struc-tural members that are subject to cyclic loading of greater than 20,000 over the life of the crane.
The
[
number of cycles for any of the crane members will be j
less than 2,000 over the life of the Reactor Building 1
l Crane. Based on this, failure due to cyclic fatigue should 17
not be of concern for this crane, and the CMAA 70 criteria for cyclic loading are satisfied.
(5)
Holstina Rope - CMAA 70 specifies a 5:1 hoisting rope
~
safety factor for the rated load plus bottom block divided -
by the number ef parts of rope. For the Reactor Building Crane, the resulting safety factor for the main hoist is:
bottom block = 16,000 lbs.
rate:1 load = 370,000 lbs.
parts of rope = 16 (Ik" each) rope published brecking strength = 152,000 lbs.
resulting safety factor = 152,000/((370,000+ 16,000)/12)= 6.3:1 For the aux. hoist:
block = 6000 lbs.
rated load = 50,000 lbs.
parts of rope = 3/4" dia.
rope published b*reaking strength = (g.t.) 45,000 lbs.
resulting safety factor =
45,000/((50,000 + 6,000)/8) = 6.4 = 6.4:1 Therefore the ropes satisfy the criteria in CMAA 70.
(6)
Hoist Drum Loads - CMAA 70 specifies that drum design should consider combined crushing and bending loads; however, EOCl 61 is not as specific. The Whiting standard design practice for this crane was to consider combined crushing and bending loads.
Therefore CMAA 70 is l
satisfled.
(7)
Holst Drum Groove - CMAA 70 specifies minimum drum groove depth and drum groove pitch; EOCl 61 does not provide such specific guidance. For the Reactor Building 18
.~.
Crane, Whiting Crane criteria required the minimum groove depth to be 3/8 x rope diameter and groove pitch to be rope diameter plus I/8 inch.
The drum pitch diameter was required to be greater than 24 x the rope ;
diameter. These standard practices were later incorpo-rated into CMAA 70.
Thus the CMAA criteria are satisfied.
(8)
Holst Holdino Brakes - CMAA 70 and ANSI B30.2 require that holding brakes have minimum torque ratings (relative to motor torque) of 125% if used with control braking other than mechanical; and 100% if used with mechanical control braking.
For the Reactor Building Crane, eddy current control braking is used. The main hoist holding broke is 210% of motor torque, and the aux. hoist holding brake is 310% of motor torque. Thus the CMAA criteria are satisfied.
(9)
Static Controls - CMAA 70 includes various criteria for crane static controls; EOCI only addresses crane magnetic controls. Since the Reactor Building Crane uses a D.C.
magnetic control system, the CMAA 70 criteria on static controls are not applicable.
(10)- Restart Protection - CMAA 70 establishes criteria for restart protection for cranes not provided with spring-
)
return controllers or momentary contact pushbuttons; this is not addressed in EOCl 61. The pendant controller for the Reactor Building Crane uses push-buttonswitches that are spring return. type, therefore, CMAA criteria cre satisfied.
DC-83-25 19
d PRC Requests (frorn draft Franklin TER)
TMI-l does not comply with the criteria.of Guideline 4.
The licensee should identify all special lifting devices associated with the heavy loads listed in Table 2.1, verify that they satisfy the guidelines of ANSI N14.6, and verify.that the stress design is based on static and dynamic loadings.
Response
Of the heavy loads listed in Table 2.1 for the Reactor Building of the Franklin TER revised Table 2.1 in Enclosure 2, only the spent fuel shipping cask and resin casks would require use of special lifting devices. Since these lifting devices are associated with a cask, and are not owned by GPUN, design evaluations of these lifting devices could not be performed. To assure that ANSI N14.6-1978 criteria are satisfied for cask handling operations at TM! in the. Reactor Building, changes are being incorporated into the " Reactor Building Crane Operation" procedure requiring that these devices satisfy the design, inspection and test requirements of ANSI N14.6-1978, including static plus dynamic ioads, prior to use of these devices for cask handling operations at TMl-l. With these changes, the FRC recommendation and NUREG guidelines will be satisfied for cask handling operations.
Detailed design specifications are not available for the special lifting devices used in the TMI-I Reactor Building and accordingly point-by-point comparisons to the applicable design sections of ANSI N14.6-1978 could not be performed. In lieu of this, stress analyses were performed for these special lifting devices. The stress analyses in conjunction with ongoing inspection and maintenance are judged to provide an adequate level of confidence in the reliability and integrity cf these devices, and provide a sufficient alternative to performance of a point-j by-point comparison to the design criteria of ANSI N14.6. The special lifting devices in the TMI-I Reactor Building considered in the stress analyses are:
i f-1)
Head and internals handling fixture with extension (reference Figure I) 2)
Turnbuckle pendants and head lifting pendants (cables)
(reference Figure 2) i DC-83-2S -
2o
3)
Internals handling odaptors, pendants, and spreader ring (reference Figure 3) 4)
Internals indexing fixture pendants (reference Figure 4)
The stress analyses were perfcrmed to the following criteria from ANSI N14.6-
-1978:
"3.2 Design Criteria 3.2.1 Stress Design Facters 3.2.1.1 The load bearing members of a special lifting device shall be capable of lifting 3 times the combined weight of the shipping container with which it will be used, plus the weight of intervening components of the special lifting device, without generating a designed shear stress or maximize tensile stress at any point in the device in excess of the corresponding minimum yield strength of their materials of construction.
They shall also be cepable of lifting 5 times that weight without exceeding the ultimate strength of the materials. Some materials have yield strength very close to their ultimate strength. When materials that have yield strengths above 80%
of their ultimate strength are used, each case requires special-consideration, and the foregoing stress design factors do not apply.
Designs shall be on the basis of the materials fracture toughness, and the designer shall establish the criteria."
Additionally, the analyses were performed to incorporate the following criterion from NUREG-0612, Section 5.1.1 (4):
"In addition, the stress design factors stated in Section 3.2.l.1 of ANSI N14.6 should be based on the combined maximum static and dynamic loads that could be imparted on the handling device, based on the chorocteristics of the crane which will be used.
This is in lieu of the guideline in Section 3.2.l.1 of ANSI N-14.6, which bases the stress design factor on only the weight (static lood) of the load and of the intervening components of the special handling device."
DC-83-25 21
Based on these analyses it was found that all members and welds satisfied the design safety factor criteria of ANSI N14.6-1978.
i.
9 Y
I i
22
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ri I
se
-l 1
I t i
i n
i p
e 4
l f I e,
.k n
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=2
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li i
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("
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g 5
1 4,a 2
j 23
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1
/-
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s I
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("
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= a
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l 7%5f=
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- J
=
/
8 lf$"? LIE?
DETAL d
FIGURE 2 TURPSUCKLE PENDANTS APO HEADLIFTING PENDANT (CABLES) 24
~~
..p.._
I e
~.
O 2a KAD APO INTERNALS HANDLING FlXTURE 8
=
l
(
]=
/:
E\\
N N
h 269 PCAD APO INTERNALS HAPOLING FIXTURE ll ll ij NTERNALS HAPOLING j
SPREADER RING I
ADAPTER APO PEPOANTS
,Y.
, g,:7 2n
- jh t
f
, g
).
jl PLENUM A55EMBLY D
f"'#'
l 252 WT 100,000 LB5 2%
l
=l C.[].3-----( U 3 2s, kh "A
j 8:
.D C. ]~: -
s-j If
/l l/
/
a mTAs.
FIGURE 3 INTERNALS HAMX.ING ADAPTER, PEIOANTS APO SPREADER RING
' v %,, p %
t 1
l O,
/
\\
[
0 N..]
=
m N
tm
==
AX 8
=
a ))
^'
[ Z1 o
4 L.
y==
sE O
LATCHNc MECHANISM 9
A W5 r
V r itn.uat.s.asuva g
q p
U nunn4ts ree.xnc w
nxru e m e urs sse
.v V7 -min:rx=
3 37 cum svuo a seit auc 1
c p-r
~
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+.
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FIGURE 4 INTERNALS HA>OLING FIXTURE PDDANTS 26
ENCLOSURE 2 RESPONSES TO REQUESTS FOR lif0RMATION IN SECTIONS 2.2,2.3, Ato 2.4 OF titC LETTER DATED 6/26/80 OVERVIEW OF NUREG-0612 EVALUATIONS GPUN has identified the following fixed handling systems at Three Mile Island (TMI-l) Nuclear Station to which NUREG-0612 is applicable. They are:
Handlino System Location Capacity (l)
Reactor Building RB - 431' el.
185 ton (M)
Crane (2)
Fuel Handling Building FHB - 380' el.
ll0 ton (M)
Crane 15 ton (A) 5 ton (Monorail) 3 ton (Monorail)
Figures I o'nd 2 show the general arrangement of the Reactor Building and Fuel Handling Building areas.
To provide on assessment of load drop consequences for heavy loods handled by these crones, the Fuel Handling Building and Reactor Building were subdivided into food impoet regions. The Fuel Handling Building was subdivided into 3 load Impact regions based, in part, on the configuration of the building and specific locations where the various heavy loads are typically handled. Figures 3 through l
10 show the subdivision of these buildings into load impact regions for evaluation l
purposes.
GPUN's first report to the NRC contains a listing of the heavy loads that are l
handled by the Fuel Handling Building and Reactor Building cranes.
The attached Table 2.1 is a revision of the table listed in our February 17, 1981 submittal, and restated in the 10/21/81 FRC TER.
27
4 REVISED TABLE 2.1 -
TMI-I FEAVY LOADS (FROM TABLE 2.1 OF FRAfELIN TER)
Weight Heavy Load 1/
(Tons) i Fuel Handling Building:
Spent Fuel Shipping Cask 152/
Pool Divider Gate 2.5 Cask Pit Gate 6
New Fuel Shipping Containers 3.5 - 4 Failed Fuel Container
.25 Fuel Transfer Corriage 1.5 Resin Cask 12 Decon Pit Cover 2.5 New Fuel Storage Pit Cover 2
Reactor Building:
Reactor Vessel Head 157 Plenum Assembly 62 ISI Tools 1-3 Reactor Coolant Pumps 35 Pressurizer Missile Shield 5.5 CRD Missile Shields 26 1/ Our response of February 17,1983 included in the list of heavy loods for the TMI-I Fuel Handling Building certain " missile shields."
These were incorrectly listed in the Table for the Fuel Handling Building, and have been deleted as noted above.
2_/ This was the food listing in Table 2.1 from the Franklin TER based on information previously submitteds by GPUN. The Technical Specifications at that time allowed handling of the cask up to 15 tons without restrictions on lift height or food path. However, the Technical Specifications also included a restriction prohibiting handling of a spent fuel shipping cask in the Unit i Fuel Handling Building. The bases section for this specification l
(Technical Specificaiton 3.11.7) states that the purpose of this prohibition is to preclude caskhandling operations at fMI-l. Concerns noted at that time l
have now been addressed and are reflected in Enclosure 2 of this response.
Based on the evaluations and analyses that have been performed as described in Enclosure 2, and proposed plant changes and modifications, this
. restriction should be removed and, accordingly, a cask of up to l 10 tons may l
be handled within the TMI-I Fuel Handling Building provided proper cautions and food paths are followed as noted in Enclosure 2.
t 28 i
Region A (Spent Fuel Pools) only requires the handling of the cask pit gate, the fuel pool divider gate, and the spent fuel shipping cask as heavy loads to be handled in this region.
Previous evaluations of cask handling operations by GPUN have demonstrated on adequate mechonism to preclude dropping or tipping of the spent fuel shipping cask into the spent fuel pools. The evaluations os described in Enclosure 2 considered handling of the two gates, and certain additional cosk drop scenarios.
Region B (area west of fuel pools) was further divided into 4 regions. The heavy loods that could potentially be handled in these regions are the fuel pool divioer gate and the cask pit gate while being moved from their in-place locations to their storage locations. Additionally, certain miscellaneous equipment will be handled in these regions while being moved to the decontamination pit for temporary storage. These loads would primarily be handled in regions B-3 or B-
- 4.
New fuel storage pit and decontamination pit covers will be handled in regions B-2 and B-3. No specific heavy loads required to be moved into E-l were identified.
Region C (Loading Bay) was also subdivided into 3 regions. Heavy loods handled in this crea include spent fuel shipping casks, as well as miscellaneous smaller heavy loods associated with Unit 2 recovery operations. In order to provide
- bounding assessments for this wide range of heavy loads,- analyses were performed using a l10 ton cask to define on acceptable region for handling of this cask. For those locations within Region C where consequences of a drop of the ll0 ton cask would be unocceptable, limiting arolyses were performed to define on occeptable handling area.
Region D is the Reactor Building and deck level south of the D rings (Reference Figure 6). This area was subdivided into regions D-l (east half) and D-2 (south I~
half). Region D-2 is located southwest of the. loop B, D ring. No heavy loads were identified that were required to be moved into region D-2. Nonetheless, systems evoluotions were performed to determine the potential consequences of load drops that penetrated the operating deck floor.
Region E (Reference Figure 7) is the operating deck level north of the D rings.
29 r*t v-* - -
-+-+w--
- +-ww r-e-w-+--eu.---e----
~----e-=+wrs mn+--
-e e
e*
--~w,
--s-+--e+
wr----e--+--
-~*-*-e*
+e-r~
t I
in this region, several heavy loods are handled including the control rod drive missile shields, the pressurizer missile shields, the reactor vessel head, and reactor coolant pumps and motors. For this region, system evoluotions were I
performed to determine the potential consequences of loods that penetrated the operating deck level within the region. The system evoluotions also considered components located at lower elevations within this region.
Region F is the loop A, D ring; and similarly region H is the loop B, D ring.
These are shown on Figures 8 and 10. Heavy loads handled within region F include the control rod drive missile shields, the pressurizer missile shields, reactor cooling pump motors, and miscellaneous smaller loads such as steam generator and piping support.
Heavy loads handled within region H include reactor coolant pumps motors plus miscellaneous smaller loods. To address the potential consequences of load drops within these regions,' systems evaluations were performed to determine the potential consequences, and the assure the capability to maintain shutdown and core cooling.
Region G is the reactor cavity area including the transfer canal (Reference Figure 9). This region wc4 further subdivided into region G-1, the crea north of the reactor center line; region G-2, the area south of the reactor cer,ter line but within the D ring; region G-3, the transfer canal area; and region G-4, tha reactor vessel. Heavy loods handled within these regions include the control rod missile shields, the reactor vessel head, and the reactor plenum assembly.
Systems evaluations were performed the regions G-1, G-2, and G-3 to assure the capability to maintain shutdown and core cooling. Structural analyses were performed for lood drops of the reactor vessel head and the plenum assembly to demonstrate the capability to withstand such load drops without loss of vessel integrity nr significant fuel domoge.
Tables I and 2 relates the defined load impact regions with the four NUREG 0612 criteria listed above. As indicated in the table, the majority of the regions were defined to assist with and focis safe shutdown evaluations to oddress NUREG-0612 Criterion IV. A combination of systems, structural, criticality and dose.
l evoluotions have been utilized to address the NUREG-0612 guidelines for TMI-l.
The specific approach chosen for each region was based on the completeness of the available information and a oreliminary assessment of the likelihood of 30
~ - - - - -. -
success of the possible appronches. Tables I and 2 also indicate the types and combinations of evaluations employed for each region for the Fuel Handling Building.
i.
EVALUATION METHODS Evaluations have been performed for each of the Fuel Handling Building and Reactor Building load impact regions to determine conformance with the
.NUREG-0612 evaluation criterio. The evaluations performed generally fall into the following four categories:
1.
Systems Evaluations 2.
Structural Analyses 3.
Offsite Dose Analyses 4.
Criticality Assessments The following summarize the evaluations performed.
Systems Evaluations As noted above, systems evaluations were performed for the Fuel Handling Building and Reactor Building load impact regions to evoluote potential load drop consequences. The objective of these systems evaluations was to determine whether required safe shutdown functions could be accomplished assuming that certain equipment was made inoperable os a result of a postulated food drop.
The systems evaluations involved several basic steps, namely:
Define the safety functions which must be occomplished for the various reactor operating modes; 2.
Identify the systems and required support systems to be relied on to accomplish each safety function; h
3.
For each load impact region, determine which components of these systems could be offected by a heavy lood drop 1
within the region; and 4.
Perform on analysis of the effects of failure of those components on the ability to accomplish the applicable safety functions.
31 l
These evaluations included identification of cabling in trays or conduit, piping, volves, and other components located in the impact region, and extended to lower elevations. Credit for In'O mning floors was not taken unless justified on 1
the basis of a structural analysis or other considerations. Loss of " component
- function was assumed if that component could be impacted. Further details on the systems evoluotions periormed and the results are contained in the response to item 2.4.
Structural Analyses As indicated in Table I o number of lood impact regions were addressed using structural analyses / evaluations. The Fuel Handling Building food drop scenarios addressed or supported with structural analyses included:
1.
Drops onto the rail / truck boy floor (Regions Cl, C2 and C3);
2.
Drops onto the environmental barrier / platform area north of the rail / truck boy; 3.
Drops onto the walls on either side of the rail / truck boy; 4.
Drops into the decontamination pit; and 5.
Drops into the cask loading pit.
Reactor Building load drop scenarios addressed or supported with structural analyses include a drop of the vessel head onto the reactor vesset, and a drop of the reactor plenum assembly back into the reactor.
The intent of the approach used was to opply worst cose loads and conditions in order to define o set of parameters that if met in food handling would not restrict the weight of loads handled in the region. Where this was not possible, l
Ilmiting loads for the various regions were defined.
l i
The results and conclusions of the structural analyses are described as appropriate in the responses to Request for Information items 2.2, 2.3, and 2.4.
The steps in tha structura! evaluation opprooch and general methodology are os follows:
32 1
- r,
,-..-,--...,4p
-m,
,--ww--.~
--,,w+
.2.-
. ~. - - - - - - - - - - - - - - - -
1.-
Identification of heavy load, handling systems, and hand-ling locations including a full characterization of the load weight, dimensions, material properties, and stru.ctural characteristics.
i 2.
Development of postulated drop scenarios based upon realistic consideration of plant procedures.
3.
Review of important structural engineering aspects of impacted structural elements to fully characterize behovlor.
For reinforced concrete and steel elements, identify drops which control " local" response (e.g.,
penetration, scabbing, spalling, perforation, etc.); loads that control "overall" structural response (e.g., large inelastic deformations or obrupt failure of principal struc-tural members, etc.); and/or foods that may induce behavior that exhibits combined response such that either overall or local failure modes would control.
4.
Incorporating I through 3 above, provide early input to the systems evaluations to factor structural information into systems evaluations assumptions.
5.
Conduct detailed structural evaluations that include:
a.
Specification of impact energy considering, as appropriate, the energy dissipated due to the trans-fer of momentum, etc.;
b.
Model develoment for assessing dynamic response utilizing empirical data os necessory; c.
Development of failure criteria based upon stability or leak tightness considerations; d.
Computation of the strain energy absorbed prior to reaching the prescribed performance limits; e.
Assessment of structural response and structural consequences of drop.
i l
The structural evaluation methodology and criteria generally follow the I
recommendations mode by the American Society of Civil Engineers Technical j
Committee on impulse and Impact Loads (Reference 1). These recommendations ore supplemented by a large body of experimental and analytical information l
-which is ' documented in reports that have been published by government, university, and industry organizations.
33
The evaluation methodology and criteria addressed below consider the two potential modes of structural behavior, local ' effects and overall structural response, respectively.
i.
Local Impact Response Evaluation:
Local impact response may lead to severe damage such as crushing, perforation, and concrete ejection in the vicinity of the impactive lood; however, overall dynumic response of the structure in the form of reactions away from the load are insignificant. The complex nature of local impact response of reinforced concrete requires evaluation using empirical formulae that are experimentally derived. The modified National Defense Research Committee (NDRC) formula (Reference 2) was chosen because it has been shown to give the best fit with available experimental data (References 3 and 4). The NDRC formulae for the i
depth of penetration, x (inches), of a solid cylindrical missile are given by:
1.8,l/2 for n 4 2.0 (1) x=
4 KNWd I d-i or
- 1.8 x=KNW V
+d for 5 >j 2.0 (2)
,,1000d.
d where W = weight of the missile (pounds) d = diameter of missile (inches)
V. = impact velocity of missile (feet /second)
N = missile shape factor
= 0.72 flat-nosed missiles
= 0.84 blunt-nosed missiles
= 1.00 spherical-nosed missiles
= 1.14 sharp-nosed missiles K = concrete penetrability factor
= 180/M(f'c = concrete compressive strength in pounds / square inch) 34
v The thickness of reinforced concrete needed to resist impact without perforation and scabbing are given by the following Army Corps of Engineers formulae which con be used in conjunction with equations I and 2 (Reference 5).
i
= 2.12 + 1.36 ( for 0.65 f f 4. Il.75 (3) g = l.32 + l.24 [ d) for 1.35 4 5 413.5 t
n (4) 1
\\
d where i = concrete thickness required to prevent scabbing s
tp = concrete thickness required to prevent perforation Equatiorc 3 and 4 were later extrapolated for small values of x/d (Reference 6)
- giving, 2
- 5.06{ h t,7,91 for 6 0.65 (5) 2 h - 0.718 t = 3.19 f
._. 3 5 (6)
(
l A 10 percent margin on thickness has been applied in the use of equations 3 through 6 as recommended in Reference I, except for concrete sections backed by steel decking where the equations were used directly.
t The effects of shape and deformability have been conservatively accounted for In' the case of the NDRC formula by odjusting the missile shape factor, N, ond/or using "equivaient" diameters.
Overall Structural Response Evaluation:
Overall structural response results from the dynamic interaction of the impact-ive ioad and the structure which it impacts.
35
~-#.-,-..-------,,w.,
v.
m.,,..
,,,,..--..r..
m m---r
..,m.-u._,
.,w.-.--7.
s.
The resultant complex forcing function produces in-structure dynamic reactions in the forms of forces, moments, and shears at points away from the impactive lood. As a rule, this forcing function is unknown; however, occasionally it can be estimated by inccrporating knowledge of the chorocteristics of the dr'opped food (weight, size, shape, deformobility), characteristics of the impacted structure (material properties, structural configuration), and the impact conditions (velo-city, orientation).
The following discussion addresses the use of energy balance methods for the evaluation of reinforced concrete structures. These techniques do not require explicit knowledge of the forcing function.
The load drop methodology incorporates the conservation of energy and momentum to calculate the transmitted kinetic energy and maximum displacement to investigate the important modes of overoll reinforced concrete behovlor.
The ob.iective of this methodology is to characterize structural behavior in terms of the available strain energy up to prescribed performance limits. These limits are dictated by either ductile or brittle modes of failure.
The ductile mode is characterized by large inelastic deflections without complete collapse, while the brittle mode may result in partial failure or total collapse. The ovalloble internal strain energy that con be absorbed by the concrete floor system without reaching those limits of unocceptable behavior is balanced against the externally applied energy resulting from a heavy lood drop.
It has been assumed that momentum is conserved, and the kinetic energy of the drop dives the mass of the floor and induces strain.
As on additional conservatism, no credit has been taken for potential sources of energy dissipo-tion through local deformation in the forms of concrete crushing and penetro-tion.
l Generally, the ultimate food of a concrete slab or beam system is reached prior to exceeding the hinge rotational capacity of particular sections provided that on j
unstable mechanism has not formed. The hinge rotational capacity was used as a criteria to set the maximum allowable level of deflection for the concrete slab or beam svstem. The hinge rotational cor acity for concrete structures was 3
36
developed in References 7 and 8 based on test results given in References 9 and 10 and is given as:
ru = 0.0065 (d/c) < 0.07 (7) where r = rotational capacity of plastic hinge (radians) u d = distance from the compression face to the tensile reinforcement c = distance from the compression face to the neutral axis at ultimate strength The maximum deflection for a concrete slab or beam with a p!astic hinge at its center is then given by:
Xm = (r l/4)
(8) u
- where, Xm = maximum deflection L = span of beam Rotations of the magnitude governed by equation 7 result in cracking vehich is confined to a region below (above) the tensile reinforcement.
Generally speaking, the section will remain intact with no crushing, spalling, or scabbing due to flexure; however, secbbing may occur as a result of shock wave motion associated with the reflection of tensile waves from the rear surface or shear plug formation. It has been conservatively assumed that scabbing does occur.
The load / deflection history up to the point of the ultimate loading, coupled with the maximum allowable deflection, defines the maximum level of strain energy obsorption provided that a shear failure has not occurred. The shear stress at limiting sections was checked and compared to allowables as specified in Chapter 11 of ACI 318-77 (Reference i I).
37
REACTOR BUILDING ANALYSES Structural analyses were performed for the head drop and the plenum assembly drop.
The maximum response of structural steel elements is determined using the commonly applied energy balance method (References I,12, and 13) by equating the externally applied kinetic energy to the available internal strain energy. Tiw maximum permissible deflection of each structural element is given in terms of an allowable ductility ratio which is defined as:
Um
- W l'
where Um = maximum permissible deflection Uy = deflection at the effective yield limit The allowable structural steel ductility ratios for impact loads have been taken from Reference I and are as follows:
ALLOWABLE MODE OF RESPONSE DUCTILITY RATIO
- l. Flexure
- open sections 12.5 i
- closed sections 20
- 2. Shear S
4 14 x 10 e 10
- 3. Compression F (KL_)2
~
y r
- 4. Torsion 0.5 Eu Ey l
l 38
. ~ _., - _. _,, _ - - -..... _ _. _ - - _.. _ - _ _ _
where Fy = minimum yleid stress of the steel K = theoretical effective length factor for compression member L = length of compression member r = rodius of gyration of cross-section Eu = ultimate strain Ey = yield strain The effective yield limit corresponds to the inflection point of an equivalent elasto-pimtic resistance displacement curve as defined by Newmark (Reference 25). For simplicity, on equivalent elasto-plastic resistance displacement curve was developed by setting the maximum resistance equal to the actual minimum yield resistance. This procedure is conservative because it neglects the strain energy associated with the strain hardening mechanism.
l l
l l
l l
l l
I 39
Discussion of Structural Margins:
In addition to the conservatisms previously mentioned, the following conserva-tisms are also inherent in the methodology used in the evaluation:
l.
Static material strengths for concrete and steel are used.
Test data shows that this property increases with the increased strain rates associated with dynamic loodings.
For example, References 13 and 15 recommend dynamic Increase factors of 1.25 for the compressive atrength of concrete and 1.20 for the flexural, tensile, and compres-sive strength of structural steel.
~2.
Equation 7 for hinge rotational copocity is used. This corresponds to rotations of the order of 2 degrees with minimum cracking and no crushing or scabbing. To meet necessary performance requirements (i.e., halting propa-goting failures), larger rotations in the range of 5 to 12 degrees could be tolerated. Such rotations would lead to crushing, spalling, and scabbing of the sectica (Reference 15); however, overall load carrying capability is expected to remain unchanged.
Experimental observations (Reference 18) suggest even further capability for well-designed and well-anchored slabs. Failure modes at such levels initially appear to be controlled by yielding in shear and flexure followed by membrane stretching until failure occurs, normally at the support edge of the slob. Use of these larger rotational copobilities would have resulted in greater energy obsorbing capabilities of the floor system.
3.
The analysis uses ACI 318-77 allowable shear stresses. A significant body of data suggests the existence of highe' shear capabilities.- (References 18 through 26).
i 4.
No credit is token for local energy dissipation associated with any crushing of the food itself or the immediate surface of the floor, as well as the steel liner in fuel pools.
5.
No credit is taken for energy dissipation associated with drog and buoyancy in water for drops inside the. spent fuel pool.
6.
No credit is token for energy dissipation by any cask impact limiter devices.
l 4o
Fuel Integrity, Offsite Dose and Criticality Considerations As indicated in Table I, the integrity of spent fuel and the associated potential offsite dose dnd criticality consequences were considered for the spent fuel pools v
(Region A) in the Fuel Handling Building. Fuel crushing analyses were performed to determine the number of assemblies that could be damaged, and thus release their contained gap activity, for load drops of the cask pit gate and the pool-divider gate.
Based on the inventory released, offsite dose analyses were performed to determine conformance with NUREG-0612 evaluation criterio.
These were performed both with and without charcoal filters to determine benefit of an ESF filter system for these scenarios. Additionally, the potential for crushing fuel in the A or B spent fuel pools due to a load drop and resulting in criticality was assessed, considering as the worst case on impact of enriched new fuel.
Based on these evaluations it was determined that the potential offsite dose and criticality consequences may not satisfy the NUREG-0612 evaluation criteria for load drops of the cask pit gate or the fuel pool divider gate. Accordingly, modifications and changes will be made as described in the response to item 2.4.1 in this Enclosure.
Fuel integrity was also evaluated as part of the plenum assembly load drop analysis.
Based on this, it was determined that the maximum urop height (dictated by the indexing fixture height) is not expected to result in fuel domoge.
41
REQUEST:
2.2 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN THE VICINITY OF FUEL STORAGE POOL NUREG-0612, Section 5.l.2, provides guidelines concerning the design and operation of food-handling systems in the vicinity of stored, spent fuel.
Information provided in response to this section should demonstrate that adequate measures have been tcken to ensure that in this oreo, either the likelihood of a lood &op which might damage spent fuel is extremely small or that the estimated consequences of such a drop will not exceed the limits sat by the evoluotion criteria of NUREG-0612, ection 5.1 Criteria I through 111.
Itom 2.2.1 Identify by name, type, capacity, and equipment designator, any crones physically capable (i.e., ignoring interlocks, moveable mechanical stops, or operating procedures) of carrying foods which could, if dropped, land or fall into the spent fuel pool.
RESPONSE
The only handling system within the TMI-I Fuel Handling Building physically capable of carrying heavy loads over or in the vicinity of the spent fuel pool is the Fuel Handling Building Crane. The crane was designed by Whiting Corporation and ir.cludes a ll0 ton main hoist and 15 ton auxiliary hoist, and 2 monorails, one of which has a 5 ton hoist and the other a 5 ton and 3 ton hoists.
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ITEM 2.2.2 Justify the exclusion of any cranes in this crea from the above category by veilfying that they are incapable of carrying heavy loads or are permanently prevented fro movement of the hook centerline closer than 15 feet to the pool boundary, or by providing a suitable analysis demonstrating that for any failure mode, no heavy food can fall into the fuel-storage pool.
RESPONSE
Since the Fuel Handling Building Crane is the only crane operating in the vicinity of the TMI-l fuel storage pools, no cranes have been excluded for this area.
4 l
43
ITEM 2.2.3 Identify any crones listed in 2.2.1, above, which you have evoluoted as having sufficient design featurs to make the likelihood of a load drop extremely small for all loads to be corried and the basis for this evoluotion (i.e., complete compliance with NUREG-0612, Section 5.l.6 or partial compliance suppiemented by suitable alternative or additional design features). For each
-crone so evoluoted, provide the food-handling-system (i.e.,
crane-lood-
. combination) information specified in Attachment 1.
RESPONSE
Due to the potential adverse consequences of a drop of a cask pit gate or a pool fue: pool divider gate into the spent fuel pools, GPUN will undertake changes and modifications to make the likelihood of a load drop extremely small for these loads. These loads will be handled using the main hoist sister hook of the Fuel Storage Building Crane, handle these gates with slinas of increased design safety factors, and upgrade lift points on these 2 gates. The Fuel Handling Building Crane operating procedure will be revised to require that 2 slings be used for hoisting of these gates, each having a design safety factor of 5 to I considering static plus dynamic loads. The cask pit gate and the pool-divider gate lifting lugs will be upgraded as required in order to provide design safety factors of 5 to I at each lug, taking into account static plus dynamic loads.
The TMl-l Fuel Storage Building Crane main hoist is provided with on upper limit switch that when the hoist block reaches a pre-determined limit of travel, will interrupt current to the hoist motor. The main hoist is also equipped with electric holding brakes and on eddy current dynamic load broke. The holding broke is solenoid released, and spring applied on loss of power to the solenoid.
The holding broke is rated at greater than 300% of the motor full load torque.
With the provisions described above, the lift points and lifting devices (s!!ngs) will be " single failure proof". The limit switch will reduce the likelihood for "two blocking" and the holding broke plus load broke will reduce the likelihood of
~
uncontrolled lowering of the load block. Additionally, use of the Fuel Storage Building Crane main hoist assures that for these foods crane components have a l
design safety factor of greater than 40 to I. Based on these proposed procedural changes and modifications, it is concluded that a drop of the cask pit gate or the pool-divider gate is of sufficiently low likelihood that they need - not be postulated.
44
NUREG-0612 requires that the load block and hook be considered as a heavy load.
The food block is used for handling several loads. In moving these loads, the hook, load block, rope, drum, sheave assembly, motor shafts, gears and other food bearing members are subjected to significant stresses approachik the load rating of the crane.
By comparison, these components are subjected to a considerably smaller load when only the hook and food block are being moved.
With the limit switch and braking features provided as noted above, and based on the relatively small food imported by the main hook and load block, it is not considered feasible to postulate a random mechanical failure of the crane load bearing components when moving the main hoist load block without a load.
S 1
t 45
ITEM 2.2.4 For cranes identified in 2.2.1, above, not categorized according to 2.2.3, demonstrate that the criteria of NUREG-0612, Section 5.1 are satisfied.
Compliance with Criterion IV will be demonstrated in response to Section 2.4 of this request. With respect to Criteria i through lli, provide a discussion of your evaluation of crane operation in the spent fuel area and your determination of compliance. This response should include the following information for each crane:
a.
Which alternatives (e.g., 2, 3, or 4) from those identified in NUREG-0612, Section 5.l.2, have been selected.
b.
If Alternative 2 or 3 is selected, discuss the crane motion limitation imposed by electrical interlocks or mechanical stops and Indicate the circumstances, if any, under which these protective devices may be bypassed or~ removed.
~
Discuss any administrative procedures invoked to ensure proper authorization of bypass or removal, and provide any related or proposed technical specification (operational and' surveillance) provided to ensure the operability of such electrical Interlocks or mechanical stops.
c.
Where reliance is placed on crane operational limitations with respect to the time of the storage. of certain quantities of spent fuel of specific post-irradiation decay
- times, provide present and/or proposed technical specifications and discuss administrative or physical controls provided to ensure that these assumptions remain valid.
d.
Where reliance is placed on the physical location of -
l specific fuel modules at c' rtain post-irradiation decay e
- times, provide present and/or proposed technical specifications and discuss administrative or physical controls provided to ensure that these assumptions remain valid.
e.
Analyses performed to demonstrate compliance with Criteria i through 111 should conform to the guidelines of Justify any exception taken to these guidelines, and provide the specific information requested I
in Attachment 2, 3, or 4, as appropriate, for each analysis performed.
l
RESPONSE
The heavy loads handled over or in close proximity to the spent fuel pools are the cask pit gate, the pool-divider gate, and a spent fuel shipping cask. As noted in the response to item 2.2.3 above, a load drop of the cask pit gate or the pool-divider gate into the spent fuel pcol need not be postulated.
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l When handling a spent fuel shipping cask, on electrical interlock system is f
octivated to control the allowable regio, for movement of the cask. Figure 9.7-2 of the FSAR shows the corrent electrical interlock system which is actuateo for movement of a shipping cask and other foods in excess of 15 tons. GPUN has I
determined that it will not use the decontamination pit for cask washdown and decontamination. Washdown of the cask will take place in the loading bay for removal of road dirt; and cask decontamination will take place over the cask pit, With these changes the controlled pathway to the decontamination pit by the electrical interlock system is not reauired and will be removed. Procedures and technical specifications also require operation of this interlock system for loads in excess of 15 tons. Due to the potential adverse consequences on safe shutdown components for lood drops into Region C (note response to item 2.4.2),
procedures and technical specifications will be revised to require use of the interlock system when handling loads in excess of 3,000 lbs. in Region C and, greater than 15 tons in other Fuel Handling Building regions with the exception of fuel handling bridge maintenance. The interlock system may be bypassed for certain loads in excess of 3,000 lbs. if the potential impact is less than or equivalent to a 3,000 lb. load at a lift height of 10 feet when handled in accordance with an opproved procedure which identifies the " safe food path" evoluoted for the food and the required lift height restrictions. Loads of up to l
3,000 lbs. may be handled in Region C3, but with a maximum lift height of 10 feet.
Over the post several years, a number of submittals have been made to the NRC concerning cask handling and cask drop evoluotions for TMl-1, including a request for amendment of the TMI-l technical specifications. Based on these previous analyses, the following conclusions may be drawn relative to cask handling operations at TMI-1: the cask travel path at the elevation 348' in the vicinity of the spent fuel pool will preclude tipping of the cask into pool B (a tip of the cock would result in the cask falling across the cask pit); and a cask carry height of I foot will not cause overall structural failure at elevation 348' for o cask drop.
Further analyses ord evolvations have been performed for consideration of cask handling in the vicinity of the ' spent fuel pool.
Systems evoluotions were l
47
l performed for Region 84 to determine potential adverse effects of scabbing or spalling due to a lood drop at elevation 348'. The systems evoluotion determined that power to I of 3 makeup pumps could potentially be lost, I of 2 BWST level
(
instruments lost, and remote operation of MUV 17 also offected.
- However, l
alternate makeup pumps, BWST level Instruments, and flow paths are available should these components be domoged. Accordingly, domoge due to a cask drop i
or tip resulting in scabbing or spalling below the Region B-4 El. 348 floor would i
not cause loss of safe shutdown functions.
Structural evoluotions were further performed for o cask drop into the cask pit to ossess potential effects on the spent fuel pool B floor. The cask pit is located on a concrete column that rests directly on the building foundation. Dynamic effects due to this load drop were evoluoted to assess the potential for scabbing, spalling, or overall failure of the fuel pool floor. These onolyses showed 'that spalling or scabbing could result for some distance out adjacent to the cask pit; however, for the components that could be domoged within this range, redundant safe shutdown components would be available outside of the domoge region.
Additionally, cask liner domoge would not be expected in pool B os a result of this load drop.
Further details on the criteria and methodology for the systems evoluotion are
^ included in the response to item 2.4.2. Additionally, requirements relative to use of electrical interlocks are contained in the " Fuel Handling Building Crane O
Operation" procedure, and as such would require management and safety review
. committee opprovol' to deviate from the procedure and bypass the interlocks.
Accordingly, with the procedural and technical specification changes described f~
obove, and the analyses that have been performed for cask handling operations in the vicinity of the spent fuel pool, handling of heavy loads over or in close proximity to the spent fuel pool satisfies evoluotion Criterio I through IV of NUREG-0612.
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P4tC REQUEST:
2.3 SPECIFIC REQUIREMENTS OF OVERHEAD HANDLING SYSTEMS OPERATING IN THE CONTAINMENT NUREG-0612, Section 5.1.3, provides guidelines concerning the design and operation of food-handling systems in the vicinity of the reactor core.
Information provided in response to this section should be sufficient to demonstrate that odequate measures have been taken to ensure that in this area, either the likelihood of a lood drop which might domoge spent fuel is extremely small or that the estimated consequences of such a drop will not exceed the limits set by the evoluotion criteria of NUREG-0612, Section 5.1, Criterio I through !!!.
2.3.1.
Identify by r.ome, type, capacity, and equipment designator any cranes physically capable (i.e., taking no credit for any interlocks or operating procedures) of carrying heavy loods over the reactor vessel.
RESPONSE: The only handling system within containment physically capable of carrying heavy loads over the reactor vessel is the Reactor Building Crone. The crane was designed by Whiting Corporation and possesses a 185-ton main hoist and a 25-ton auxiliary hoist.
49
titC REQUEST:
2.3.2. Justify the exclusion of any cranes in this crea from the above category by verifying that they are incapable of carrying heavy.loods, or are permanently prevented from the movement of any lood either directly over the reactor vessel or to such a location where in the event of any lood-handling-system failure, the load may land in or on the reactor vessel.
RESPONSE: The Fuel Handling Bridges used for refueling operations are excluded. They are sized to handle single fuel assemblies, i.e., no heavy loads os defined in NUREG-06l2 are handled by these handling systems.
9 I
50
6 bEC REGUEST:
- 2.3.31(dentify any cranes, listed in 2.3.1 above which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely small for all loads to be corried and the basis for this evaluation (i.e.; complete comliance with NUREG-0612, Section 5.l.6, or partial compliance supplemented by: suitable alternative or additional design features).
For each crane so evaluated, provide the-lood-handling system (i.ey crane-lood-combination)
Information specified in Attochment 1.
RESPObs3E: Based on the design evaluation of the Reactor Building crane os described in enclosure I and the ongoing inspection and maintenance of this crone, load handling reliability with this crone is considered to be very high.
However, since this crane does not fully satisfy the criteria contained in Section S.I.6 of NUREG-0612 (i.e., " single failure proof"), the consequences of postulated load drops have been oddressed for the Reactor Building as described in response to item 2.3.4.
51 l ~
14tC REQUEST:
2.3.4 For crones identified in 2.3.1 above not categorized according to 2.3.3, demonstrate that the evolution criteria of NUREG-0612, Section 5.1, are satisfied. Compliance with Criterion IV will be demonstrated in your response to Section 2.4 of this request. With respect to Criterio I through lil,* provide a discussion of your evoluotion of crone operation in the containment and your determination of compliance.
This response should include the folowing information for each crane: -
a.
Where reliance is placed on the installation and use of electrical interlocks or mechanical stops, indicate the circumstances under which these protective devices con be removed or bypassed and the administrative procedures invoked to ensure proper authorization of such action.
. Discuss any related or proposed technical specification
~
concerning the bypassing of such interlocks.
b.
'Where reliance is placed on other, site-specific considerations (e.g.,
refueling sequencing),
provide present or proposed techni al specifications and discuss administrative or pFysical controls provided to ensure the continued validity of such considerations.
c.
Analyses performed to demonstrate compliance with Criteria i through 111 should conform with the guidelines
~
of Attachment 5.
Justify any exception taken to these guidelines and provide the specific information requested in Attachment 2, 3, or 4, as appropriate, for each analysis performed.
RESPONSE: The Reactor Building crane handles the reactor vessel head and the plenum assembly over the reactor vesse!. Additionally, the CRD missile shields are handled over the reactor vessel head. To address the potential consequences f
of dropping these various loads, analyses were performed of a drop of the plenum j
assembly onto the reactor core and a drop of the reactor vessel head onto the L
vessel flange.
In addition, an assessment was performed of the CRD missile shield drop to determine plant conditions and the capability to maintain continued decay heat removal for such a load drop. The following summarize the analyses that were performed and the results obtained.
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Plenum Assembly Drop Onto the Reactor Core The plenum asssembly, located directly above the reactor core, is removed as a single component before refueling. It weighs approximately 119,000 pounds with its lifting rig and is removed and replaced according to Procedures 1504-8 and 1506-l. The lifting system used to move the plenum includes the containment polar crane, the plenum assembly lifting rig, and various adapters, pendants, and fixtures. These items have been evoluoted and found to meet the intent of industry standard ANSI Bl4.6-1978, with the exception of certain welds, but these will be upgraded. Inspection procedures provide increased assurance that these devices would be able to maintain their design load margins.
The polar crane was evaluated to industry standards CMAA 70-1975 and ANSI B30.2-1976 and found to meet these standards.
Notwithstanding the fact that the lifting system, including the Reactor Building polar crone and associated lifting devices, complies with the intent of applicable industry standards and possesses demonstrated margins to failure, on evaluation has been performed for a postulated drop of the plenum onto the reactor core from its maximum lifted height.
The plenum assembly consists of a plenum cover, upper grid, control rod assembly (CRA), guide tube assemblies, and a flanged plenum cylinder with openings for reactor coolant outlet flow. The plenum cylinder (127-3/4" height) consists of a large cylindrical section with flanges (upper flange has 166-7/8" O.D.) on both ends to connect the cylinder to the plenum cover and the upper grid.
The plenum cover is constructed of a series of parallel flat plates 53
Intersecting to form square lattices; it has a perforated top plate and on integral flange at its periphery. The cover assembly is attached to the plenum cylinder top flange. The perforated top plate has matching holes to position the upper end of the CRA tubes. Lifting lugs are provided for remote handling of the plenum assembly. These lifting lugs are welded to the cover grid. The upper grid consists of a perforated plate which locates the lower end of the individual CRA guide tube assembly relative to the upper end of a corresponding fuel assembly. The CRA guide assemblies provide structural attachment of the grid assembly to the plenum cover by welding to the plenum cover top plate and bolting to the upper grid.
Locating keyways in the pienem assembly cover flange engage the reactor vessel flange locating keys to align the plenum assembly with the reactor vessel, the reactor closure head CRD penetrations, and the core support assembly. The bottom of the plenum assembly is guided by the inside surface of the lower flange of the core support shield.
During removal and replacement of the plenum, clignment is accomplished using the internals indexing fixture which is positioned on the reactor vessel flange.
The internals indexing fixture is cylindrically shaped with internal locating keys that mcte with the locating keyways in the plenum assembly cover flange.
Perfect alignment of the plenum within the internals indexing fixture is required to slide the plenum in and out of the reactor vessel; therefore, the maximum postulated drop height corresponds to the indexing fixture height which is 73th inches.
Conservatively neglecting the energy absorbing effects of drag as the plenum l
travels through water and a " dashpot" effect that exists due to the close tolerance of the plenum within the core barrel, the kinetic energy of the drop is 729,000 foot-pounds.
The total kinetic energy has conservatively been assumed to reach the core and directly load the fuel assemblies, although certain physical limitations exist and potential additional sources of energy disspitation have conservatively not been l
54 l
l c.
Included. The impact lood is transmitted uniformly from the plenum upper grid to the fuel assembly upper end fittings through the 16 control rod guide tubes, and to the fuel assembly lower end fittings. The fuel rods are not significantly loaded unless the upper end fittings are driven into the fuel rods due to deformation of the guide tubes through buckling. The energy absorbed by the guide tubes failing in on inelastic buckling mode has been conservatively ignored.
An analysis was performed of the fuel pin energy absorption that is allowable.
Individual fuel rods are predicted to buckle elastically between spacer grids at a Euler critical buckling load (Per) of 210 pounds. Strain energy con be absorbed beyond the point of reaching (Per) through bending until the fuel cladding has reached 1% strain or yielding in 50% of the section. These criteria are based upon Irrodioted properties of Zircoloy-4, the cladding material. This evaluation determined that the total strain energy that can be absorbed into the fuel in the core is expected to be greater than the energy input due to the load drop.
l Based upon this evaluation, in the unlikely event that the Reactor Building crane and its associated lifting devices fail while the plenum is at the maximum point of carry at which'it can impact the core, we conclude that the fuel cladding is not expected to rupture and radioactive gases are not expected to be released.
Reactor Vessel Head Drop Onto the Reactor Vessel The reactor pressure vessel (RPV) head is hemispherically shaped and weighs approximately 310,000 pounds with the control rod drive (CRD) service structure and the RPV head lifting rig. The RPV head has an I.D. n 152.4 inches and a flange O.D. = 200 inches, giving a flange area of approximately 13,180 square l
inches.
The RPV head is removed and reptemd according to procedures 1504-7, " Closure Head Removal," and 15062,"Closuie Head installation". The head lifting system l
Is essentially identical to the plenum assembly lifting system, with the exception of different pendants. Similarly, the lifting system complies with the intent of l'
industry standards as previously described.
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The head is moved laterally along a direct path between the reactor vessel and I
the head storage stand, which is located at the north end of the operating floor.
These procedures allow the head to be removed without simultaneously ' raising the refueling canal water level if radiation levels permit. The heoki may be lifted from the fienge elevation of 32! ft. to above elevation 346 ft while directly over the reactor vessel flange. Calculations of a vessel head drop from ten feet above the flange have determined that vessel skirt failure cannot be precluded for such a load drop. Accordingly further calculations were performed using finite element techniques to determine capability of the vessel skirt for a head drop of 45 inches (height required to clear the head guide studs).
Two postulated RPV head drop cases were considered. In the first case the impact lood to the RPV fiange was assumed to be axisymmetric and the drop height 45 inches, in the second case the RPV head was assumed to drop in such a way that the RPV flange is loaded over a very small portion of its perimeter by a
" point lood". This would correspond to the head hitting the flange after tilting 19 degrees due to a failure of a single pin, lug, or arm in the lift rig. The dissipation of drop energy associated with the impact of a mass to another mass (Ref. 35) was accounted for using conservative assumptions for the mass distribution of the RPV and the support skirt. The coefficient expressing tha ratio of available energy after the impact to the kinetic energy of the dropping body before impact was found to be 0.408 in the axisymmetric case. The impact load path in the RPV is from the flange through the RPV shell to the support skirt which is bolted to a concrete slab.
In the analysis of the RPV it is conservatively assumed that the point where the skirt is attached to the concrete slab is fixed, i.e. flexibility under this point is not accounted for.
The RPV shell and the support skirt have almost equal material-centerline diameters (179.4 inches for the shell and 177.5 inches for the skirt), thus the RPV l
and support skirt can be approximated for a simplified calculation by a cylinder with constant diameter and varying thickness.
To determine the load Py corresponding to first occurrence of yield, spring constant of the structure under the force acting at the flange, and the energy associated with the elastic deformations of the structure a simplified hand calculation was made for the axisymmetric case where the load applied at the RPV flange was assumed to be 56
~
~
n carried by only membrane forces in the RPV shell and support skirt.
To investigate the significance of bending effects at the RPV shell-support skirt Joint a finite element analysis (using the SAP V computer code) was performed.
The (static) force P octing at the RPV flange that causes yield at thd skirt was y
found to be 69,257 kips when membrane forces only were considered, and 51,903 kips when bending effects were included and yielding occurs only at the outermost fiber. The reduction in the force P due to bending effects was thus y
. 25%
For the.nonaxisymmetric case a simplified calculation was performed again 4
neglecting bending effects in the RPV shell and support skirt plate.
It was assumed that in the support skirt and the lower part of the RPV cylinder vertical normal stress varies os cos (9), the total expression for this stress being determined from the force and moment equilibrium conditions for on entire cross section of a cylinder. Also in this case the governing factor is yielding in the skirt as load is increased. As in the axisymmetric case, it was assumed that the bending effects reduce the force corresponding to first occurrence of yield in the skirt by 25% in determination of the spring constant for the nonoxisymmetric case, for the uppermost section of the RPV cylinder, where the stress does not vary as cos (@), the spring constant of the axisymmetric cose was conservatively used.
i or both the axisymmetric and nonaxisymmetric case, the ductility requirements in order to accommodate the drop energy are within required criteria listed in the lead-in sections to Enclosure 2. Since the auctility requirements are lower than the allowable ductility ratios, it is concluded that the RPV structure con l
sustain the loads associated with both RPV head drop scenarios (i.e. oxisym-metric and nonoxisymmetric drops from just above the head guide studs) without l
loss of Integrity.
1 Procedures for remcving and installation of the vessel head will be revised to require that the vessel head not be lifted greater than 45 inches above the vessel l
i l
57 l
- -,. ~. -. - - - - _. - -. - -
CRD Missile Shields The CRD missile shields will not be handled above the reactor vessel until the rods are fully inserted and the reactor is in cold shutdown conditions. At this point, the reactor is on decoy heat removal and crushing of CRD housings would not have on adverse effect on core cooling. Loading of the control rods is unlikely due to the significant crushing of the housings required before the end of the control rod drive shafts are impacted (with control rods inserted).
i 0
.O 58
MIC REQUEST:
2.4 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN PLANT AREAS CONTAINING EQUIPMENT REQUIRED FOR REACTOR SHUTDOWN, CORE DECAY HEAT REMOVAL, OR SPENT FUEL POOL COOLING NUREG-0612, Section S.I.S, provides guidelines concerning the design and operation of load handling systems in the vicinity of equipment or components required for safe reactor shutdown and decay heat removal.
Information provided in response to this section should be sufficient to demonstrate that adequate measures have been taken to ensure that in these areas, either the likelihood of a load drop which might prevent safe reactor shutdown or prohibit continued decay heat removal is extremely small, or that damage to such equipment from load drops will be limited in order not to result in the loss of these safety-related functions. Cranes which must be evaluated in this section have been previously identified in your response to 2.1-1, and their loads in your response to 2.1-3-c.
1 2.4.1 Identify any cranes listed in 2.1-1, above, which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely small for all loads to be carried and the basis for this evaluation (i.e., complete compliance with NUREG-0612, Section S.I.6, or partial compliance supplemented by suitable alternative or additional design features). For each crane so evaluated, provide the load handling system (i.e.,
crane load combination) information specified in Attachment 1.
RESPONSE
The Fuel Handling Building Crane has not been determined to fully meet the criteria of Section S.I.6 of NUREG-0612; however, as noted in the response to item 2.2.1 above, handling of the cask pit gate and the fuel pool divider gate by the Fuel Handling Building Crane is judged to be sufficiently j
j reliable, provided certain changes and modifications are made, such that a drop f
of these loads need not be postulated. The required changes and modifications I
include use of the main hoist for handling these loads, and increased design safety factors for slings and lifting lugs providing at least a 5 to I design safety factcn... static plus dynamic loads should one sling or lift point fail. On this basis, these loads may be excluded from further structural and systems evaluations described below.
59
ITEM 2.4.2 For any cranes identified in 2.1-1 not designated as single-failure-proof in 2.4-1, a comprehensive hazard evaluation should be provided which includes the following information:
a.
The presentatiori in a metrix format of all heavy loads orld; potential impact areas where damage might occur to -
safety-related equipment.
Heavy loads identification should include designation and weight or cross-reference to information provided in 2.1-3-c. Impact areas should be identified by construction zones and elevations or by some other method such that the impact area can be located on the plant general arrangement drawings.
Figure i provides a typical matrix.
b.
For each interaction identified, indicate which of the load and impact area combinations can be eliminated because of separation and redundancy of safety-related equipment, mechanical stops and/or electrical interlocks, or other site-specific considerations. Elimination on the basis of the aforementioned considerations should be supplemented by the following specific information:
(l)
For load / target combinations eliminated because of separation and redundancy of safety-related equipment, discuss the basis for determining that load drops will not affect continued system operation (i.e., the ability of the system to perform its safety-related function).
(2)
Where mechanical stops or electrical interlocks are to be provided, present details showing the areas where crane travel will be prohibited. Additionally, provide a discussion concerning the procedures that are to be used for authorizing the bypassing of interlocks or removal stops, for verifying that interlocks are functional prior to crane use, and for verifying that interlocks are restored to operability offer operations which require bypassing have been completed.
(3)
Where load / target combinations are eliminated on the basis of other, site-specific considerations (e.g.,
maintenance sequencing), provide present and/or proposed technical specifications and discuss administrative procedures or physical constraints invoked to ensure the continued validity of such considerations.
60
' ITEM 2.4.2 (Continued)
For interactions not eliminated by the analysis of 2.4-2-b, c.
above, identify any handling systems for specific loads -
which you have evaluated as having sufficient design ~
features to make the likelihood of a load drop extremely small and the basis for this evaluation (i.e., complete compliance with NUREG-0612, Section 5.l.6, or partial compliance supplemented by suitable alternative or odditional design features). For each crane so evaluated, provide the food handling system (i.e.,
crone lood combination) Information specified in Attachment 1.
d.
For interactions not eliminated in 2.4-2-b or 2.4-2-c, above, demonstrate using appropriate analysis that damage would'not preclude operation of sufficient equipment to allow the system to perform its safety function following a 4
load drop (NUREG-0612, Section 5.1, Criterion IV). For each analysis so conducted, the following information should be provided.
r (1)
An indication of whether or not, for the specific food being investigated, the overhead crane handling system is designed and constructed such that the hoisting system will retain its food in the event of seismic accelerations equivalent to those of a safe shutdown earthquake (SSE).
(2)
The basis for any exceptions taken to the onolytical guidelines of Attachment 5.
(3)
The information requested in Attachment 4.
RESPONSE
P FUEL HAMX.ING BUILDING: Safe shutdown components are located below the spent fuel pool (Region A), areas west of the spent fuel pools (Region Bl,82, B3 and 84) and the rail / loading bay areas (Regions Cl, C2 and C3). A combination cf structural and systems analyses were performed for load drops in each of these regions. The types of evoluotions performed for each defined load impact region are listed in Table 1. Structural analyses were performed following the methodology described in the introductory sections to this enclosure, and included the following structural analyses:
o drops onto the rail - truck boy floor (Region Cl);
o drops onto the environmental barrier / platform area north of the rail - truck boy (Region Cl);
61 i
1 l
l drops onto the walls on either side of the rail / truck bay o
(Regions Cl and C2);
~
L o
drops in Region C3; drops into the decontamination pit (Region B2); and o
drops into the cask loading pit (Region A).
o s
The following provides a summary of the methodology applied for performance of the systems evoluotions of the TMI-I Fuel Handling Building.
Systems Evaluation Methodoloay As part of the evaluation of heavy load handling operations at TMI-l a number of potential load impoet regions were identified and addressed by performing
" systems evaluations." The regions involved are Load Impact Regions A, B and C in the Fuel Handling Building. Figures 3, 4 and 5 identify these Load Impact Regions on a 348' el. floor plan of the Fuel Handling Building. Note that Regions B ond C have been div ded into subregions.. All regions and subregions extend vertically downward to the 281' el. slob, unless justified otherwise, e.g., by structural analysis.
The objective of the " systems evoluotions" was to demonstrate that safe shutdown and long term cooling could be achieved and maintained assuming that certain combinations of equipment were lost due to a postulated food drop. This demonstration of safe shutdown capability was for the purpose of achieving compilance with Criterion IV of NUREG-0612, Section S.I.I.
In order to demonstrate the ability to safely shutdown and cool the core, it was necessary to (l) identify the safety functions to be accomplished, (2) identify the plant systems required to occomplish these functions, (3) identify the equipment that could potentially be lost if a lood drop were to occur in certain plant ureos (the Lood impoet Regions), and (4) determine the resultant effects of the loss of this equipment on the safety functions required to achieve safe shutdown.
62
To determine the functions that must be accomplished to achieve safe shutdown, it was assumed for all regions investigated in the Fuel Handling Building that the reactor was at 100% power at the time the food drop was postulated. The system functions required to be accomplished are those necessary to achieve and maintain cold shutdown. Only " safety systems" were relied on in accordance with the criterio of NUREG-0612.
~
The systems and functions required to accomplish safe shutdown utilized for the purpose of evaluating the potential consequences of postulated heavy load drops are identified below.
SAFETY FUNCTION SYSTEM (1)
Short Term Reactivity Control RPS CRDCS CRA (2)
Lono Term Reactivity Control Makeup (HPI) utilizing BWST as Boron Source
-(3)
Intermediate Cooling System (IC) or Seal Injection (Makeup /HPI)
(4)
RCS Pressure Control PZR Heaters or Makeup (HPI),
(to assure natural circulation)
ADV's (S)
Secondary System Inventory Emergency Feedwater System utilizing CST as Water Sourco (6)
RCS Inventory (Level Control)
Makeup (HPI) utilizing BWST as Water Source I
(7)
Decay Heat R,emoval j
Short Term OTSG's (Natural Circulation in RCS)
ADV's j
Long Term Decay Heat Removal System (8)
Event Monitorina See Table C-2 (9)
Air Handlina Rx Building and Control Building Emergency Ventilation Systems Intermediate Building Ventilation 63 i
~.
. ~
SAFETY FUNCTION SYSTEM (10) Auxiliaries Electrical Power & Control Batteries, Chargers & Inverters (Supports Safety Functions I-9)
Diesel Generators arid Support l
Systems See Table C-2 for Electrical Distribution
- Panels, SWGR, Control Centers and Circuits Instrument Air Compressors Makeup (HPI) Support Systems Makeup Pump Main or Auxiliary (Support Safety Functions 2,3 & 6)
Lube Oil Systems Nuclear Services Closed Cooling Water System (NS)
Nuclear Services River Water System (NR)
Decay Heat Closed Cooling Water System (DC)
Decay Heat River Water System (DR)
Decay Heat Removal Support Systems - DC (Supports Safety Function 7)
DR Air Handling Support Systems (Supports Safety Function 9)
Control Room NS NR Reactor Building-r Rx Building Emergency Cooling River Water System (RR)
Intermediate Cooling System NS Support Systems NR (Supports Safety Function 3)
Eoch region of concern was evaluated to determine which of the system components of interest could be lost.as a result of a load drop. The evaluation was performed based on plant drawings, FSAR system descriptions, the Appendix R Report, equipment location drawings, piping layout drawings, conduit and cable tray layout drawings, and piping and instrumentation drawings, as well as a walkdown of plant areas.
An assessment wcs then performed of the consequences of losing the components in the region with regard to occomplishment of the system function, and the overall consequences 64
considering the combinations of potential component failures in different
- systems, i
l In performing the systems evaluations, equipment was assumed lost based on a l
conservative judgement regarding the impacted area in a region (area of l
Influence). The area of influence for a region was either assumed to be broader than the actual area of impact could be (e.g. on entire Load impact Region at all elevations) or was constrained to a smaller area of influence (if necessary) by considering the characteristics of the load, the building structure and the results of structural analyses.
Several important general and specific assumptions regarding equipment loss are listed below:
(1)
With respect to system mechanical components (e.q.
pipes, valves), it was assumed that if such components were in a region of concern then they were lost, unless they could be shown to be safe from domoge based on other considerations such as distance from actual load impact point.
(2)' With regard to pipe breaks, the offected system or subsystem was assumed lost unless the capability to isolate, repair or plug the break could be demonstrated.
(3)
Credit was given for certain manual actions, such as operating valves, in the same manner as was relied on by GPUN for the Appendix R evaluations.
Results of Evaluations Based on the structural and systems evaluations, the regions where potential problems exist are R(.gions B-1, B-2, B-3 and C-3.
The specific concerns and.
proposed corrective measur'es for each of these regions are addressed below.
Load drops for the fuel pools (Region A) are discussed in the response to item 2.2.4.
65
Region B-l:
The problem for Region B-1 is the potential for damaging safe shutdown piping in the pipe and ventilation chose that extends from the 348' el. down to the 281' el.
If the plant were at power or at cold shutdown when such a heavy load drop occurred, successful accomplishment of the following safety functions can not be assured:
(1)
Long Term Reactivity Control (2)
RCS inventory Control (3)
RCS Integrity (RCP Seal Failure)
(4)
Long Term Decay Heat Removal (5)' Control Room Emergency Ventilation The safe shutdown systems that could potenticily be lost are the intermediate Cooling System, the Nuclear Services Closed Cycle Cooling System and the Decay Heat Closed Cycle Cooling System.
In addition to the potential for breach of piping in critical systems, certain safe shutdown cabling could also be impacted in Region B-!. However the affected system functions could be accomplished by redundant components powered and controlled by cables that are not offected by a load drop in this region.
i l
The " Fuel Handling Building Crane Operation" procedure will be revised to prohibit heavy load movements in the northwest corner of the operating deck defined by on area encompassing Region B-l.
l l
Region B-2:
t Region B-2 (see Figure 4) of the 348' el. contains the Nuclear Services Closed Cycle Cooling and Intermediate Closed Cycle Cooling Surge tanks and the new fuel storage pit. These tanks and the suction lines from the tanks to their associated pumps could be directly impacted by a heavy load drop into this region and if breached could potentially result in loss of both of these systems.
l 66
1 in addition, both Decay Heat Closed Cycle Cooling Surge Tonks are located in Region B-2 Immediately below the 348' el. The safety functions that could potentially be lost are the same os those listed above for Region B-l.
. One additional concern in this region is that the power cables for all three makeup pumps are located in the region. However, these cables are located below the 305' elevation. It is extremely unlikely that a heavy load drop into this region from above the 348' el. would have any effect on this cabling. The only feasible way the cabling could potentially be impacted is if a heavy load were dropped onto the 348' elevation floor from sufficient height to cause either overall failure or perforation of the floor with subsequent cascoding failures of the floors below. The only heavy loods identified as being handled in this region are the cask pit gate and the new fuel storage pit cover; odditionally the fuel pool divider gate would be handled in close proximity to this region. As noted in response to item 2.4.1, changes and modifications for handling of these gates will improve lood handling reliability such that a drop of these gates need not be postulated. The new fuel storage pit cover only needs to be raised sufficiently
(_10 inches) to be placed on top of the decontamination pit cover. Due to the 4
weight of this cover and the small lift height, damage to safe shutdown components below this region is not expected to occur due to a load drop. Based on this, failure of the makeup pump cables at the 281' el is not considered as o potential result of a postulated heavy load drop in Region B-2.
i The " Fuel Handling Building Crone Operation" procedure will also be revised to prohibit heavy food movements within Region B-2 (see Figure 4), other thon:
movement of the new fuel storage pit cover (Region B-2) at the minimum height r
necessary to place it over the decontamination pit (Region B-3); or movement of the cask pit or fuel pool divider gates to their storage locations; or movement of the decontamination pit cover at the minimum height necessary to place it over the new fuel storoge pit cover.
Region B-3:
Region B-3 (see Figure 4) contains the decontamination pit and a portion of floor area at Elevation 348 ft. The decontamination pit will not be used for cask washdown or decontamination; however, it is used for handling of smaller loods, r-67
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some of which may be considered as heavy loads. These loods may include o
various contaminated equipment or tools associated with operations. Nuclear Services Closed Cycle Cooling System piping is located immediately below the decontamination pit floor. This piping is the inlet and outlet piping for both
_ Spent Fuel Pool Coolers. The volves available to isolate the coolers from the remainder of the Nuclear Services Closed Cycle Cooling System are also located in this region. Accordingly, a breach of the piping upstream or downstream of E
the inlet or outlet isolation volves respectively, could result in on unisoloble break in the Nuclear Services Closed Cycle Cocling System. The potential consequences of this' system loss is an inability to accomplish the Control Room 4
Emergency Ventilation function, or to provide cooling to Letdown Coolers.
Structural analyses were performed in accordance with the methodology described in the introductory sections of this Enclusure to determine the maximum load that may be handled within the decontamination pit, such that a lood drop would not result in scabbing or spalling. Thest analyses determined that a load of 3 tons could be acceptably handled within the decontamination pit.
Accordingly, the " Fuel Handling Building Crone Operation" procedure will be revised to prohibit movement of any heavy load in excess of 3 tons over the decontamination pit boundaries other than the decontorninction pit cover and new fuel pit cover.
Region 5-4:
Region B-4 (see Figure 4) encompasses a floor area at elevation 348' odjacent to pool. B. The only heavy loods identified as having to be handled within Region B-4 ore a spent fuel shipping cask and the cask pit gate. As noted in the response to item 2.4.1 above, changes in procedures for cask pit gate handling will provide lood handling reliability such that a drop of this gate need not be postulated.
Potential consequences for o drop or tipping of a shipping cask within this region l-are described in the response to item 2.2.4. Based on the evoluotions described h
in the response to 2.2.4, a load drop in Region B-4 will not cause loss of safe shutdown capability, i
l l
68
Region C-l:
Region C-1 (see Figure 5), contains, below the 305' elevation slab and the railway boy floor, cabling associated with I train of Nuclear Services, Decay Heat Services and Reactor Building Emergency Cooling River Water Systems. Cobling associated with the redundant train of these pumps is not located within Region C-l. Accordingly, damage to redundant safe shutdown components within Region C-1 will not result in loss of safe shutdown capability. A load drop within Region C-1 could cause resulting concrete scabbing or spalling in adjacent Region C-3; this concern is discussed in the Region C-3 description below. A cask drop analysis was performed to determine the potential for a cask drop from the 348' elevation onto t'he wall below the 305' elevation slab separating the north loading bay area from the rail boy to determine the potential for damaging redundant safe shutdown cables that pass through this wall within Region C-3. The analysis was performed for a cask drop within Region C-l; the results of this analysis determined that redundant safe shutdown cables in C-3 would not be offected.
Region C-2:
Region. C-2 (see Figure 5), contains, below the 305' elevation slab, cabling associated with I train of pumps in the Nuclear Services, Decay Heat Services, and Reactor Duilding Emergency Cooling River Water Systems. Cabling for the redundant train of these pumps is not located within Region C-2. Accordingly, a load drop Ir. Region C-2 would not' cause loss of safe shutdown capability.
. Structural analyses were also performed of the wall below the 305' elevation slab between Regions C-1 and C-2 to determine the potential for a drop on this wall to cause loss of redundant cabling on either sides of this wall. These analyses showed that even for a drop of a 110 ton cask from the 348' elevation within Region C-2, damage to the floor slab and to the wall would not be sufficient to cause loss of redundant shutdown cabling located in regions on either side of this wall.
69 d
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Region C-3:
Region C-3 (see Figure S), contains below the 305' elevation slab, cabling associated with redundant pumps in the Nuclear Services, Decay Heat Services and Reactor Building Emergency Cooling River Water Systems. Loss of these cables could potentially result in loss of the capability to accomplish the following safety functions:
(I)
Long Term Reactivity Control (2)
RCS Inventory Control (3)
RCS Integrity (RCP Seal Failure)
(4)
Long Term Decov Heat Removal (5)
Reactor Building and Control Room Emergency Ventilation Structural analyses were performed for a drop of a 110 ton cask from the 348' elevation within Region C-l.to determine the potential for subsequent effects, such as scabbing or spalling, or for overall floor failure within Region C-3.
Bowd on these analyses, it was determined that if the electrical interlock system were mddified such that the eastward limit for cask handling, in Region C-l were moved further to the west (i.e.,17'6" from the centerline of the east crone rail), and load handling be allowed within Region C-2, tne consequences of a cask drop wovid be limited 3uch that redundant safe shutdown cabling within Region C-3 would not be domoged. Based on this, modifications will be made to the fuel hcndling building electrical interlock system. The present electrical interlock system is shown in Figure 9'.7-2 of the FSAR. The ottoched Figure 12 illustrates the proposed revised limits for the electrical interlock system. This Includes removal of the travel path to the decontamination pit at elevation 348'.
Additionally, structural analyses have determined that loods of up to 3,000 lbs.
may be handled in Region C-3 with a maximum lift height of 10 feet.
l l
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70
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Procedures and technical specifications will be revised to require use of the interlock system when handling loads in excess of 3,000 lbs. in Region C, and greater than 15 tons in other Fuel Handlinq Building regions with the exception of fuel hand lIng bridge maintenance. The interlock system may be bypassed for certain loads in excess of 3,000 lbs. If the potential impact is less than or equivalent to a 3,000 lbs. lood at a lift height of 10 feet when handled in accordance with an approved procedure which identifies the " safe load path" evaluated for the load and the required lift height restrictions. Loads of up to 3,000 lbs. may be handled in Region C3, but with a maximum lift height of 10 feet.
Results and Conclusions - Fuel Handlino Building Based on the systems and structural evaluations that have been performed for Regions B and C, certain changes to procedures and modifications to equipment will be required as described in the preceding response. With these changes and modifications, the potential consequences of a heavy load drop will not result in loss of safe shutdown capability. Since cask handling operations and associated movements of cask pit gate, or fuel pool divider gate are not anticipated for some time, and with measures implemented related to crane inspection and testing, crane operator training and qualification, crane design evaluation, load handling procedures, and sling selection, sufficient assurance is provided that load handling will be performed safely within the fuel handling building in the interim period until the modifications and changes described in this response are l
completed.
l l
Additionally, with these changes and modifications, and based on evaluations and analyses described in this submittal, the Technical Specification limitations on I
cask handling, pool gate movements, and handling of heavy loads in the vicinity of the spent fuel pools may be removed (Technical Specifications 3.11.6, 3.11.7 and 3.II.8).
l 71
REACTOR BUILDING Similar systems evaluations to those for the Fuel Handling Building were performed to evaluate the consequences of postulated food drops in the Reactor Building. However, the evaluations for the.
Reactor Building differed from those performed in the Fuel Handling '-
Building for the following reasons:
1 (1)
-Load drops were only postulated to occur in the Reactor Building when the plant was in the cold shutdown condition and on Decay Heat Removal System cooling.
Load drops were postulated to occur in the Fuel Handling Building during any plant condition including 100% power operation.
(2)
Load drops in the Reactor Building in certain arcos have the pntential for breaching the Reactor Coolant System pressure boundcry.
Load drops in the Fuel Handling Building did not have this potential.
(3)
Load drops in the Reactor Building cor. occur with the reactor in various configurations (e.g. RV head on, RV head off, RV head on, but defensioned). This was also possible for the Fuel Handling Building, but because of the possibility of direct impact of the RCS and DHR systems, is more significant for the Reactor Building.
As a result of these differences, the safety functions to be accomplished, the systems required to respond to the drop consequences and the evaluation approach are somewhat different as described in the following sections.
I f
h 1
1' 72
Plant Conditions As noted above, the initial plant condition for all drops in the Reactor Building was the cold shutdown condition with reactor decay heat being removed by the Decay Heat Removal (DHR) System in the shutdown cooling mode.
This assumption was utilized 'because none of the Reactor Building heavy loods identified in GPUN's first heavy loads report to the NRC (February 17,1981) are required to be moved in the Reactor Building during other plant operating modes.
In addition, various RCS configurations are possible at the time of the postulated food drop. Emergency Procedure EP1202-35, Revision 9 dated August 10, 1983 entitled, " Loss of Decay Heat Removal System" was utilized to identify the spectrum of possible initial RCS configurations that could be encountered.
These initial RCS configurations are (I)
RV head on and bolted, RCS intact (2)
RV head on, but defensioned (3)
RCS open below the 321' el.
(4)
RV seal plate and transfer canal drain flange installed, RV hecd off (5)
RV seal plate not installed, RV head off As described in the following sections, depending on the initial plant i
configuration, the avoilable alternative core cooling schemes to DHR cooling con
_ vary.
Note that with the initial plant condition being the cold shutdown condition, significantly more time is potentially available to respond to the food drop l
consequences and accomplish mitigating octions. With a loss of DHR cooling, for l_
cxample, in certain situations and depending on the time after shutdown, there can be many hours available to restore core cooling before sufficient boil off occurs to uncover the core. Accordingly, the potential consequences of a heavy 1
73
J load drop during cold shutdown are inherently less significant in terms of plant
' safety, as compared to the potential consequences of a heavy load drop during power operations.
~
(
Safety Functions and Minimum Required Systems / Components There are two basic load drop outcomes of safety significance related to heavy load drops in the Reactor Building during cold shutdown conditions. These are:
(1)
Loss of DHR cooling; and (2)
A breach of the RCS piping that could result in the inability to maintain DHR cooling.
Accordingly, the safety functions to be accomplished following load drops in the Reactor Building are:
(1)
With No RCS Break - Maintain and/or restore decay heat removal with DHR or other systems; and (2)
With an RCS Break - Maintain core coverage and occomplish long term core cooling following the break.
As indicated above, the systems capable of accomplishing these two safety functions con vary depending on the plant configuration at the time of the drop.
Emergency Procedure 1202-3S, Revision 9 covers response to these two basic drop consequences for various plant configurations. Utilizing this procedure and cssociated referenced procedures, a minimum set of systems / components was identified to accomplish the safety functions defined above for various postulated load drop scenarios. The postulated load drop scenarios for various load impact regions and the required components for each scenario were identified in the course of the evaluation as described in the following section.
}
i 74
-~
V Reactor Buildina Systems Evoluotion Approach The load impact regions defined for the Reactor Building are identified in Figures 2-6 through 2-10. They were defined in such a way as to bound the potential area of damage for any particular heavy load drop, based on architectural considerations and the areas over which heavy load movements occur. Systems evolvations were conducted of all load impact regions identified in these figures with the exception of Region G-4, Reactor Vessel. This region is addressed separately in response to item 2.3.
Emergency Procedure 1202-3S, Revision 9 was the basic reference utilized to define the evoluotion logic. Using this procedure and related reference material, on evoluotion flow chart was developed that illustrates the paths to acceptable (NUREG-0612, Criterion IV Met) and unacceptable (Potential Problem) evaluation conclusions for postulated load drops within the various load impact regions.
This flow chart is shown in Figure i1. The hexagon shaped flow chart elements define the initial evaluations for each load impact region, including the potential consequences (e.g., DHR drop line failure, RCS break) of the postulated locd drop within the region. Depending on the answers to these questions for a region, other parts of the flow chart come into play. The diamond shaped clements in these other parts of the flow chart identify the minimum required systems / components located inside the Reactor Buildina that are required in order to proceed successfully along the various flow chart paths. As can be seen from a review of the flow chart, the components required are almost exclusively piping runs for various sytems. This is because shutdown components requiring ciectrical power, e.g., pumps, are located outside of the Reactor Building.
i l
Composite piping plans and tracing of cabling in trays and conduits was performed to determine the components that could potentially be damaged due to a lood drop within each region. Based on the components that could be t
domoged within each region, the evaluation verified that a success path from Figure iI is available utilizing components that would not be damaged due to the lood ' drop.
1
{
75 i
i
4 Results and Conclusion - Reactor Buildino A demonstration of compliance with NUREG-0612 was possible for all load impact regions in the Reactor Building with the exception of Region D-2.
Region D-2 encompasses the Reactor Building piping penetration region and therefore redundant piping runs for critical systems are included in this region.
The principal piping runs of interest within the region are those associated with the DHR system (dropline and both injection lines) and the HPI System (all 4 injection lines). On this basis, it was concluded that drops of heavy loads within this region should be avoided. Based on this, the Reactor Building polar crane operating procedure will be revised to prohibit heavy loads from being handled in Region D-2.
1
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REFERENCES 1.
Civil Engineerino and Nuclear Power, Report of the ASCE Cornmittee on Impactive and Impulsive Loads, Vol. V, American Society of Civil Engineers, September 1980 2.
Effects of impact and Explosion, Summary Technical Report of Division 2, National Defense Research Committee, Vol. I, Washington, D.C.,1946 3
Vossallo, F. A., Missile impoet Testing of Reinforced Concrete Panels, HC-5609-D-I, Colspan Corporation, January 1975 4.
Stephenson, A. E., " Full Scale Tornado Missile impact Tests," Electric Power Research Institute, Final Report NP-440, July 1977 5.-
Beth, R. A. and Stipe, J. G., " Penetration and Explosion Tests on Concrete Slobs," CPAB Interim Report No. 20, January 1943 6.
Beth, R.
A.,
" Concrete Penetration," OSRD-4856, National Defense Research Committee Report A-319, March 1945 7.
ACI 349-76, Code Requirements for Nuclear Safety-Related Concrete
_S_tructures, Appendix C "Special Provisions for Impulse and Impactive Effects," American Concrete institute,1976 l
l 8.
Kennedy, R. P., "A Review of Procedures for the Analysis and Design of l
Concrete Structures to Resist Missile impact Effects," Journal of Nuclear Encineering and Design, Vol. 37, No. 2, May 1976 l
~
9.
Mottock, A. H., " Rotational Capacity of Hinging Region in Reinforced Concrete Beams," Flexural Mechanics of Reinforced Concrete, ASCE 1965-50 (ACP SP-12), American Society of Civil Engineers,1965 l
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Corley, W. G., " Rotational Capacity of Reinforced Concrete Beams,"
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-1 j
11.
Building Code Requirements for Reinforced Concrete, ACI 318-77, American Concrete Institute, December 1977 12.
Structural Analysis and Desian of Nuclear Plant Facilities, American
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" Design of Structures fo.- Missile impact," Topical Report BC-TOP-9A, Bechtel Power Corporation, September 1974 14.
" Development of Criteria for Seismic Review of Selected Nuclear Power Plants," NUREC/CR-0098, N. M. Newmark and W. J. Hall for the U.S.
Nuclear Regulatory Commissicn, May 1978 15.
Structures to Resist the Effects of Accidental Explosions, TM5-1300, Department of the Army, Washington, D.C., July 1965 16.
Desian of Structures to Resist the Effects of Atomic Weapons - Strenath of Materials and Structural Elements, TM5-856-2, Department of the Army, Washington, D.C., August 1965 17.
Personal communication between Professor William J. Hall and Howard A.
Levin, October 5,1981 18.
Wang, C. K. and Salmon, C. G., Reinforced Concrete Desian, Intext i
Educational Publishers, New York,1973 1
19.
Ferguson, P. M., Reinforced Concrete Fundamentals, J. Wiley, New York, 1973 78
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Untrauer, R. E. and C. P. Siess, " Strength and Behavior in Flexure of Deep Reinforced Concrete Beams Under Static and Dynamic Loading," Civil Engineering Studies Structural Research Series Report No. 230, University of Illinois, Urbano, October 1961
(
21.
Austin, W. J., et al, "An Investigation of the Behavior of Deep Members of Reinforced Concrete and Steel," Civil Engineering Studies Structural Research Series No.187, University of Illinois, Urbana, January 1960 22.
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32.
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Roark, R. J., and W. C. Young, " Formulas for Stress and Strain," Fifth Edition, McGraw-Hill,1975 80
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TABLEI APPLICABLE EVALUATION CRITERIA FOR EACH IMPACT REGION IMPACT REGIONS FUEL HANDLING BUILDING REACTOR BUILDING l
CRITERIA 1 A
Bl B2 B3 B4 Cl C2 C3 D
E F
G H
I (Radiological) xII 1/
J/
1/
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1/
11 II 1/
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1/
h (Criticality)
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x 2d/
x 2d/
1/
1/
1/
1/
1/
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x2/
x 2d/
x 2/
X
.l/
Radiological and criticolity considerations may not satisfy regulatory criteria for a drop of the cask pit or canal goles into the pool. Accordingly modifications and changes are proposed to improve handling reliability for these foods so that such a drop need not be postulated.
2/
With snodifications and changes to be made for the cask pit and canal gates,no heavy foods were identified where o drop irwa.fing the pool liner need be postulated.
3/
Systems Evoluotions 4/
Struetural Evoluotlens 5/
To be addressed in o subsewnt submittol.
TABLE 2 APPLICABLE EVALUATION CRITERIA FOR EACH IMPACT REGION d
kMP^C REACTOR BUILDING EG O CRITERIA DI
'D2 E
F Gl G2 G3 G4 H
lIl l
(Rodiological) l l
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II Fuel domoge due to plenum assembly drop would not occur os described in Appendix A and Appendi 1'/ Criticolity assessment is discussed in Section 2.3.4.
j.
El Systems Evoluotions in Appendix C.
SI Structural Evoluotions in Appendix 8.