ML20023A808
| ML20023A808 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 02/26/1982 |
| From: | Bayne J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20023A799 | List: |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612 JPN-82-25, NUDOCS 8210200039 | |
| Download: ML20023A808 (65) | |
Text
r pv j POWER AUTHORITY OF THE STATE OF NEW YORK to coLUMSUS Ctactc Ncw Yoh. N. Y. IOo19
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GEORGE L. ING A-LS vtCE C=asawa.
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February 26, 1982 L T.*4^.E ci,
JPN-S2-25 U.~ El "
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- 'irector of Nuclear Reactor. Regulation U.
S.
Nuclear Regulatorv Commission tra s hi ng t on,
D.
C.
20555
'ttention:
Mr. Domenic Vassallo, Chief Operating Reactors Branch No. 2 Division of Licensing
. eject:
James A.
FitzPatrick Nuclear Power Plant Docket No. 50-333 Control of Heavy Loads - NUREG-061:
- .eferences:
- 1..
- Letter, D. G.
Eisenhut ( ':RC ) tc-all Operating Reactors dated December 22, 1960 2.
Letter, J.
P.
Bayne (PASNY) to T.
A.
Ippolito (NRC) dated October 15, 198. (JPN-81-82).
ear Sir:
Reference 1 requested a review of heavy load-handling operations and a two-phase submittal of evaluations of their conformance to the guide'ines of NUREG-0612.
The Power Authority completed the first phase-of this review and submitted the six month report via Reference 2.
This report identified procedure changes necessary to meet the NUREG-0612 interim action guidelines.
The Authority cmnitted to implement these changes during the Reload 4/ Cycle 5 refueling outage now in progress.
In accordance with this commitment, these procedure changes have been completed.
The enclosed nine month report provides the results of the second phase of the review and the Power Authority's response to the items in Sections 2.2, 2.3 and 2.4 of Enclosure 3 to the December 22, 1980 letter.
The analyses completed to date and summarized in this report indicate that the consequences of certain load drops do not, or might not, meet the guidelines of NUREG-0612.
The Power Authority will prohibit the load lifts identified below until further evaluation demonstrates that the likelihood of the drop is sufficiently small or, constraints will be imposed on the lift so'that the consequences of a drop are ac6eptable.
'N 8210200039 P
-...--------------.-----_---.------_------.--_----,_--a.---_--_
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1.
The steam separator assembly will not be lifted into a: out of the reactor vessel.
2.
The refueling slot shield plugs will not be lifted unless both of the following conditions are met:
a.
no freshly discharged spent fuel ir ntored in the vicinity of the lift; and, b.
no fuel is stored in non-boral racks in the vicinity of the life.
3.~
No spent fuel or radioactive waste shipping casks will be lifted in the reactcr building.
4.
Recirculation pump motors will not be lifted in the ncrthwest equipment hatch (Region 3) unless the plant is in the refueling condition (as defined in the enclosure).
5.
No heavy loads will be lifted through the RiiR heat exchanger hatches unless the plant is in the refueli:49 condition (as defined in the enclosure).
The Power Authority will also evaluate reactor vessel head drop scenarios in addition to the one included in the enclosed report.
These evaluations, and the others mentioned above, will be submitted er soon.as they are completed.
The load handling restrictions described above will assure the safety of load handling operations in the interim.
If you have any further questions, please do no hesitate to sontact us.
Very truly yours,
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P.
Bayne Senior Vice President Nuclear Generation cc: Mr. J.
Linville Resident Inspector U. S.
Nuclear Regulatory Commission P. O. Box 136 Lycoming, New York 13093 Mr. Ron"Barton United Engineers & Constructors, Inc.
30 S.
17th Street Philadelphia, PA 19101
s RESPONSES TO REQUESTS FOR INFORMATION IN SECTIONS 2.2 AND 2.3 OF ENCLOSURE 3 TO NRC DECEMBER 22,1980 LETTER 2.2 SPECIFIC REQUIREMENT 5 FOR OVERHEAD HANDLING SYSTEMS OPERATING IN REACTOR BUILDING NUREG-0612, Section 5.1.4, provides guidelines concerning the design and operation of load-hcndling systems in the vicinity of spent fuel in the recctor vessel or in storage. Information provided in response to this section should
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demonstrate that adequate mecsures hcve been taken to ensure that, in this creo, either the likelihood of a load drop which might damage spent fuel is
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extremely smo!!, or.that the estimated consequences of such a drop will not
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exceed the limits set by the evaluation criteric of NUREC-0612, Section 5.1, Criteria i throug'h 111.
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ITEM 2.2-1 identify by name, type, capacity, and equiprnent designator, any i
crones physically capable (i.e., ignoring interlocks, moveable mechanical stops, or operating procedures) of carrying loads over spent fuel in the storage pool or in the reactor vessel.
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RESPONSE
Taree handling systems operating in the reccior building cre copcble of ccrrying loads over spent fuel in the storage pool or in the reactor vessel. These handling systems are described in Table 1.
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TABLEI HAtOLING SYSTEMS CAPABLE OF CARRYING LOADS OVER SPEND FUEL IN THE STORAGE POOL OR IN THE REACTOR VESSEL EQUIPMENT CRANE TYPE CAPACITY DESIGNATOR Recctor Building Crcne Overhead Bridge Mcin Hoist - 12S tons CR-2 Aux. Hoist - 20 tons Aux. Hoist - h ton Refueling / Service Pil!cr 7S0 lbs.
JC-25 A,B,C Jib Cranes (3)
Refueling / Service Bose 750 lbs.
JC-27A,B,C Hoists (3)+
+ The 3 refueling / service hoists are each mounted on the 3 refuel / service jib crci.;.
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s ITEM 2.2-2 Justify the exclusion of any crancs in this crea from the above rifying that they are incapable of carrying heavy loads or are permanently prevented from movement of heavy loads over stored fuel or into any location where, following any failure, such load may drop into the reactor vessel or spent fuel storage pool.
RESPONSE
The Refueling / Service Jib Cranes and the Refueling / Service Hoists mcy be excluded from the above category. Justification for exclusion of these handling systems wcs provided in our response to Section 2.1, item 2, of our initici submittal responding to the NRC letter of December 22,1980 (letter from George T. Berry to Dctrell G. Eisenhut dated October 15, 1981). That response indicated that these hoists cre being derated from 1,000 lbs. to 750 lbs. ond will be clearly merked with the lower rating. Therefore, these hoists are excluded from the NUREG-0612 criteric since they wi!! not be allowed to handle loads greater than 750 lbs.
3
ITEM 2.2-3 Identify any crones listed in 2.2.-1, above, which you have evaluated cs having suf ficient design features to nake the likelihood of c locd drop extremely small for c!! loads to be corried and the bcsis for this evaluatiun (i.e., co oplete comp!!.
cnce with NUREG-0Cl2, Sectico 5.1.4. or partial compliance supplemented by suitob!c citernative or additional design features).
For ecch crane so evaluated. provide the load-handling-system (i.e.,
crane-load-combination) infor notio.
specified in Attachment 1.
RESPONSE
The Reccior Building crane was evaluate.1 in industry standcrds CMAA 70-1975 (reference 5) and AN$1 B30.2-!?76 (reference 6). It was found to meet these sicndards, with two exceptions which were justified in reference 3.
Therefore, bcsed on those evaluations, the reliability of the Reactor Bucoing crane is demonstrated.
1 Notwithstanding the fact that the lif ting system including the Reacter Building 3'
i crene, lif ting slings cnd strongbacks, complies with the intent of applicable I
. industry standcrds and possesses demonstratec mergins to f ailure, rather then relying on the reliability of the lif ting system, on escluation has been performed to ossess the consequences of postulated drops of heavy loads.
Therefore, cithough these heavy load drops need not be postulated, the load handling s
reliability of the Recctor 8vilding crane was conservatively not relied on (except as indicated below), and the consequences of postulated load drops have been evolua' ted.
i The only case where load handling relicbility was considered was with respect to the main hoist load block and hook. NUREG-0612 (refereom 34) requires that the load block and hook be considered as a heavy load. The load block is used for hcndling numerous loads, including the reactor vessel head, drywell head, shield plugs, and the dryer and separator units. In moving these foods, the hook, load
- block, rope, drum, sheave assembly, motor shaf ts, gears, and other load becring members are subjected to significant stresses approaching the load rating of the crane. By comparison, these components are subjected to a considerably smaller load when only the book and load block are being moved. Based on this, it is not considered feasible to postulate a random mechenical failure of the crane load bearing components when moving the crone load block alone.
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The only feasible failure modes for dropping of the main hook and food block j
would be:
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A control system or operator error result;ng in hoisting of the block to o "two blocking" position with continued hoisting by the motor and subsequent parting of the rope (this situation con be prevented by operator action prior to "two blocking" or by on upper limit switch to terminate hoisting prior to "two blocking"); and 2)
Uncontrolled lowering of the load block due to failure of the holding brcke to function (the likelihood of this con be made small by use of redundant holding brakes).
The Fit; patrick Reactor Building crone is provided with two diverse upper limit switches to interrupt power to the hoist motor prior to "two blocking." When power is removed, holding brokes are automatically cpplied. One of the two limit switches is'c geared limit switch driven of f the drum shaf t.
The other is o counter weigi.t switch that is released when the load block comes up ogoinst a trip bar; the ' trip bor will stop power to the hoist below the low point of the sheave ossembly.
The holding brokes are solenoid released, and spring opplied on loss of power to the solenoid. Two holding brakes are provided, either of which hos sufficient capacity to hold the roted load (ecch broke is 150% of f all motor torque).
Additionally, inspection and maintenance procedures assure that the limit switches and holding brakes are functional and properly odjusted.
With the provisions described above, the two diverse limit switches will reduce the likelihood for "two blocking" and the two holding brakes will reduce the likelihood of uncontrolled lowering of the load block. Based on these features, it is concluded that a drop of the load block and hook is of sufficiently low likelihood that it does not require load drop onalyses.
Nonetheless, on analysis of a load block and hook drop from the highest possible carry height onto the operating / refueling floor was performed to verify the capability of the floor to withstand the impact of such a drop. The results of the.
onclysis indicate that while concrete scabbing on the underside of the floor is 5
predicted, no gross failure or penetration wi!I result. The consequences of scabbing have been considered in the systems evaluations and weee found to be occeptable.
Therefore, although drop of the load block and book need not be postulated, even if they were to drop on the operating / refueling floor, the consequences are acceptable.
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m ITEM 2.2-4 For crones identified in 2.2.-1, obove, not coiegorized according to 2.2-3, demonstrate that the criteric of NUREC-0612, Section 5.1, cre satisfied.
Compliance with Criterion IV will be demonstrated in response to Section 2.3 of this request. With respect to Criterio I through Ill, provide a discussion of your evoluotion of crane operction in the Reactor Building and your determination of complinnce.
ITEM 2.2-40 Where reliance is placed on the installation and use of electri-cc! interlocks or mechanical stops, indicate the circumstances under which these protective devices can be removed or bypassed and the cdministrative procedures invoked to ensure proper cuthorization of such action.
Discuss any related er proposed technical specificctions concern:ng the bypass of such inter!ocks.
RESPONSE
An interloc'< system is provided for the Re ctor Bui: ding Crcne that prohibits movement of aeovy loods over the Spent Fue! %ol caring normd focd handling operations. This interlock system is described in the response to item 3.c in our initial submittal responding to the NRC's letter of December 22.
1980.
The keys which allow operatica in modes other than the NORMAL mode must be obtained from the plant Shif t Supervisor. Bypouing of any of the interlocks modes is controlled by plant procedures.
Use of these procedures requires management approval and use of a Work Tracking Form (WTF).
The Shif t Supervisor must cpprove ecch WTF prior to commencing work. These procedures cre contained in Maintencnce Procedure MP 17.1. Deviations from procedures require opprovcl in the menner described in the response to item 3.b in our previous submittal.
Notwithstanding the fact that the interlock system prohibits movement of heavy loads over the Spent Fuel Pool, in the unlikely event that the inter!vek iciis to protect against drops of certain items handled near the pool (e.g., the portable radiation shield and the refueling slot plugs), procedures will be implemented to assure that no fuel is stored in nonboral rocks near the edge of the pool and that no newly spent fuel will be stored in that crea. Therefore, the consequences of even unlikely drops into the spent fuel pool are determined to acceptably comply with the guidelines of NRC NUREG-0612.
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One of the interlock modes (the Cask Hcndling Mode) described in that response does allow movement of heavy locds such as casks over a small l' x 4' crea in the southwest corner of the pool, i.e., load movement is prohibited directly over spent fuel. Concerns regarcing cask tipping cnd ccsk impact on the pool floor have previously been raised by the NRC and addressed of the Authority. To cddress these concerns, we described, in our letter dated Nove.nber 12,1974, a Fuel Cask Drop Protection System proposed for installation at JAF. By letter dated April 22, 1977, we indicated that we would reevaluote the need for instcIlotion of thct system prior to ship:nent of spent f uel. which was not caticipctec before tne icte 1980s.
As a result of the currea.t review of heavy load handling operations at JAF, we hcve cetermined that positive protection of spent fuel in the pool is provided through use of the Feel Ccsk Drop Prctection Syste n.
Therefore, instc!Iction of on cpproved licensed system w:ll be complete prior to ship;nen+ cf spent fuel at JAF.
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ITEM 2.2-4b Wh re reliance is placed on the operation of the Stand-by Gcs Trectment System, discuss present and/or proposed technical specifications and administrative or phyiscal controls provided to ensure that these assumptions remain valid.
RESPONSE
In no cases is relicnce placed on operation of the Stand-by Gcs Trectment System. However, JAF technical specifications require that second-cry containment be maintained cnd that one train of the Stand-by Gcs Treatment System be operable when handling irradicted fuel or fuel casks.
t ITEM 2.2-4c Where reliance is placed on other site-specific considerations (e.g., refueling sequencing), provide present or proposed tech-nical specifications, and discuss admin:strative or physical controls provided to ensure the validity of such considerations.
RESPONSE
In no cases is reliance placed on other siteopecific considero-tions. However, as discussed in response to item 2.3 following, certain load lif ts
- cre scheduied to occur only during the refueling mode.
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ITEM 2.2-4d Analyscs perform d to demonstrate compliance with Criteric !
through 111 should conform to the guidelines of NUREG-0612, Appendix A.
Justify any exception taken to these guidelines, and provide the specific information requested in Attachment 2,3, or 4, as appropriate, for each onelysis performed.
RESPONSE
There are three potential consequences of interest when considering load drops onto the open reactor vessel. They cre: 1) loss of reactor vessel integrity,2) fuel cladding demoge and the resultant radiological dose, and
- 3) fuel crushing and the possibility of a resulting criticality condition. Criteric I through lli in Section 5.1 of NUREG-0612 cddress each of these potentic!
conse:;uences.
Tne evaluations below hcve been performed to cddress these issues.
The recctor pressure vessel (RPV) head shown in Figure I weighs 73 tons, including the weight of the RPV head strongback. Removci cnd reassembly of the head are accomplished according to Fitzpctrick Maintenance Procedures MP 4.! cnd MP 4.2, respectively.
During normel refueling operations, the RPV head assembly is lif ted out cf the recctor cavity from about elevation 345', to the operating / refueling floor ct elevation 369'-6".
The vessel head is lifted at a time when no water is in the cavity. As a result, the evoluction of the head drop wcs performed assuming a 25 foot drop through cir. Once et the desired heignt, the RPV head is moved south toward the head holding pedestcl which rests on the operating / refueling floor.
Reassembly is in the reverse order. Several head drop scencrics over the RPV con be postulated in the unlikely event of a failure of the reactor building crane.
The potential for fuel damage, or o loss of safe shutdown capability affecting the ability to get water to the core for cooling purposes, was reviewed for the case of the RPV head drop from the normal carry height of the reactor building crane, 25 feet through cir, impacting on the RPV flange. The general methods of analysis which are documented in references 9,10, and 29 through 31 were used, along with parameters which are applicable to the Fitzpatrick plant. The RPV head drop was analyzed using two methodologies. The behavior of the RPV below the head flange, and at the support skirt, was 'nalyzed for the load resulting from the 25 foot drop of the head, includinr consideration of load amplification due to the dynamic impect factor. Stre,ses and stability were 10
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evaluated at significant regions.
In addition to verifying the load carrying capabilities of the RPV, on assessment of the energy of the 25 foot drop of the head was performed. The objective of this evoluction was to verify that the RPV and skirt con survive the effects of the drop through energy obsorption and dissipation. The resulting velocity of impact of this 25' drop through cir, wcs ec!culated bcsed on the equation of motion:
d = V t + b ct2 i
Thus, the RPV hecd wcs found to impcci the flange at c velocity of 40.1 feet per second. The energy cf the RPV heud drop con be eciculcred, considering the velocity at impcct end recognizing that momentum is conserved when the RPV head impacts the fienge.
Therefore, assuming a reactor vessel head and essembly weight of 73 tons, the resultant energy was calculated to be 8.18 x 105 f t.-lbs.
The major portion of the impoet load of the RPV head is transmitted directly to the RPV fionge. The load path is then through the RPV shell to the supporting skirt which absorbs the entire impact.
The dynamic model conservatively neglects energy absorption by the reactor internals. An assessment of the load path and supporting system revealed the response behavior of the system and the critical load to the component due to the load drop.
The critical load wcs defined as that load which caused initial yielding of the weakest member. The system wcs then cnolyzed to determine its capability to absorb energy based on elastic response to this critical load. Thus, the energy obsorbing capacity of the system was calculated to be 1.3 x 106 ft.-lbs. Since this is greater than the energy of the head drop, the RPV and support skirt are capable of absorbing the energy of the RPV head drop.
In addition to verifying that the energy of the %od drop con be cbsorbed by the RPV and the skirt, stresses and stability at significant orecs (such as the vessel wall, the lower vessel head crec, the support skirt, and the RPV and skirt interface) were evaluated, in order to compute the stresses along the lood path through the vessel and supporting skirt, it was necessary to calculate the dynamic impact factor, IF. It was assumed that the stresses are distributed in il f
the some menner as for the case of static loading; however, they cre increased by this dyncmic impoet fcctor. According to reference 32, Roark and Young, this impact factor can be represented by the ratio:
IF = 6 i/ 6 = Wi/W = 1 + (1 2h/ !)b
- where, Sg=
vertical deformction on impact static vertical deformation S =
Wi= force or effective weight upon impact W=
static force or weight of dropped locd h
height of crop
=
The obove formula is bcsed on the assumption that impoet strair.s the e!cstic body the some way as s*ctic loading, i.e., that all of the kinetic energy of the moving body is expended in producing this stroin. Actually, on impact some of the kinetic energy is dissipated, cnd this loss, which con be calculated by equating the momentum of the er; tire system before end af ter impact, is most conveniently taken into account by multiplying the available energy (measured by h or by v2) by a factor K. The above equation con then be rewritten os:
IF =
si/5 = W;/W = 1 + (l+2Kh/ 6)b A number of approximate models are availcble for estimating the energy loss factor K (reference 32). If a moving body of mess M strikes axially one end of a bor of mass Ml, the other end of which is fixed, then I + (1/3XM /M)
I K
=
(1 + (h)(Ml/M)) 2 If there is a body of mass M2 attached to the struck end of the bar, then I + (1/3XM /M) + (M /M) i 2
K
=
(1 + (h)(Ml/M) + (M /M)) 2 2
Therefore, using the above expressions, the dynamic impoet factor con be determined.
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The stiffness of the RPV shell and the RPV support were modeled along with the associated masses of the actual system. The static deflection of the RPV flange due to the load was calculated, based on the spring stiffness of the load supporting system (the RPV shell, the bottom head region, and the support skirt region). Bcsed on this deflection and the transfer of momentum between the head and the RPV upon impact, the dynamic impact factor of the head drop wcs calculated, and the resulting dynamic load was determined. The stress was then calculated for the recctor vessel, and wcs found to be significantly below the minimum yield stress of 50,000 psi, and the ultimate stress of 75,000 psi. In the bottom becd crec, and in the supporting skirt, the stresses were also s;gnificently below code c!!owables.
Ir. addition to evaluating the stress levels in the reactor vessel and support, the stability cf the recctor vessel wcs also considered, to determine the potentici for buckling. The reccior vessel wo.; conservatively represented as c thin wnlled cylindrical tube under uniform longitudinct compression, and the thecretico!
buckling stress wcs calculated to be about 400,000 psi. Therefore, since the ecleulated axic! load of the impcet on the vessel wall is 10,800 psi, it is obvious that stability of the reactor vessel shell is not a problem. Similcrly, a review of the buckling potential of the skirt shows that the potential for buckling in this crea is not predicted, since the theoretical value is calculated to be about i65,000 psi.
Based on the evaluations above, reactor vessel integrity is maintained and no fuel damage is predicted as a result of the reactor vessel head drop.
The limiting situation for fuel demoge was judged to be the postulated drop of the upper internals package into the vessel. This includes on evaluation of both the steam dryer and the steam separator. A conservative structural evoluotion was performed to determine if fuel integrity could be demonstrated for each of these postulated drops.
In order to determine the worst case between the dryer drop of 24' through air and. e steam separator drop of 32.5' through water, the impact energies of each of these drops were determined. Based on the result of the impact energy 5
evoluotions (the separator drcp energy was determined to be 3.4 x 10 f t-lbs and 13
c>
Structural damage in Region 6 is predicted to be limited to scabbing of concrete under the refueling sicb. Dcmage to equipment at the 344' elevation below this region was therefore evaluated on the bcsis of the potential for the scabbing.
Loss of the only safe shutdown equipment that could be impacted in this region would result in the inability to utilize one of the two redundant Low Pressure Coolant injection (LPCI) loops to provide makeup during reactor depressuriza-tion / cooling. The other loop of LPCI would be unaffected. In addition, other systems such cs Low Pressure Core Spray could be utilized to accomplish this function. Therefore, the consequences of load drops in Regicn 6 were deter-mined to acceptcbly comply with NRC evaluation criteria.
Region 7 Equipment, fue! containers, and shipping ecsks cre moved in the scuthecs!
equipment hctch (Region 7). Structural evaluations were performed to ceter-mine the consequences of postulated drops in terms of overall structural failure and local structural response. cor those heavy loads whose controlling mode of response was determined to be local in Table 2, the consequences of the locd drops were found to be acceptcble. That is, no penetration of the 272' elevation floor sich was predicted for these load drops. However, the bounding drops were determined to be those controlled by overall structural modes. Specifically, the shipping casks were determined to result in the worst case consequences.
Shipping casks have not been selected to date, however the 34 ton Chem-Nuclecr Systems, Inc. (CNS) 4-4S cask is being considered as a possible condidate fortransporting radioactive material. This cylindrical cask is 1731/8" long and 42 1/2" in diameter except for 31 S/8" at the end which is 401/2" in diameter. In oddition, the consequences of handling accidents postulated for a larger, as yet unspecified fuel assembly cask weighing 110 tons, measuring S' in diameter and 18' in height were also evaluated.
The 272' elevation concrete slab at the bottom of the equipment hatch is 2'3" thick and reinforced with #11 reinforcing bars each way - each foce. The sicb is supported by a 7' 2" thick,12' wide beam that spans north - south, opproximately 24' between the inside and outside crescent walls, and approximately 32' from 37
the insida crcsetnt wall to o support p:destal which is located at the reactor side of the suppression pool. The beam is the principal load carrying member and has been provided for rail car loadings. The beam is located to the east side of the equipment hatch projection, leaving only the 2'3" slob for protection of the northwest qucdrcnt under which the suppression pool is located.
For the ecsk drop evaluation, the potentici for perforating the 2'3" sicb in the northwest quedront of the hatch was found to be high. Additionally, it was determined that the becm is subject to shear failure for postulated drops of either cosk in the immedicte vicinity north or south cf the intersection with the supporting inside wall of the crescent crec. Therefore, it wcs concluded that postu!ated drops in the equipment hotch creo could cause obrupt failure of the concrete structure cbove the supprer.sion pool leading to impcet and the pote 1tial for loss of lecktight integrity of the suppresion pool.
Potential dcmage to safe shutdown equipment in this region was investigated in the vicinity of the hatch at the 272' elevation and below.
Sofe shutdown equipment in these crecs whose loss could potentially result in the incbility to achieve and maintain safe shutdown includes: 1) the suppression pool and 2) RHR service water piping affecting both of two redundant loops. The suppression pool is required to successfully accomplish reactor pressure relief /depressurization and initial cooldown if isolated from the main condenser. RHR service water is required to prcvide cooling water to the RHR heat exchangers, which are utilized to remove decay heat in the long term in all of several different possible system arrangements for accomplishing decoy heat removal.
While the consequences of the bounding cask crop onto the track bay floor below the hatch were found to be unacceptable, several options exist and are currently being evaluated for developing a safe solution for moving casks in this region.
Results of those evaluations will be provided to the NRC at a later date. No handling of ensks in this region will be performed until a safe solution has been implemented.
38
Region 8 Region 8 includes the equipment hatch in the northwest quadrant of the reactor building.
No heavy loads currently corried in this equipment hatch were determined to cause unacceptcble structurci response for a drop onto the floor sicb ct elevation 272' for this region. However, the potential for movement of a recirculation pump motor in this region may exist in the future. In such a ccse, the structural response of the floor at 272' would be bounded by the cask drop cnolyses previously described for Region 7, i.e., failure of the floor con not be precluded.
Potential demoge to scfe shutdown equipment in this region was investigated in the vicinity of the hatch at cod below elevation 272'. The only safe shutdown equipment thot could be impacted in these creas whose loss could eventuc!iy result in on incbility to cchieve safe shutdown is the suppression pool.
The suppression pool is required to successfully cecomplish reactor pressure relief /
depressurization and initici cooldown, if isolated from the main condenser. For this reason no loads that could result in demoge to the suppression pool, if dropped, will be handled in this hatch unless the plant is shutdown cnd depressurized, and on on cppropriate mode of long term cooling. The only such load currently anticipated is a recirculation pump motor, os mentioned above.
With this restriction on load handling operations, the consequences of postulated load drops into Region 8 were determined to acceptably comply with NRC evoluction criteric.
i Regions 9E cnd 9W Regions 9E and 9W cre the RHR heat exchanger hatches on the east and west ends of the reactor building. A systems approach to evoluoting postulated load drops in these regions was performed.
Potentici domoge to safe shutdown equipment in these regions was investigcted separately in the vicinity of each of the hatch openings at the 326' elevation and the 300' elevation, and within and below the RHR heat exchanger cubicles at the 272' elevation. Evoluotion was not required at the 344" elevation, because this crea was included as part of the evoluotion of Region 4.
Investigations et the intermediate reoctor building l
39 g
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clavations wtro undtrtaken because of the potential for o large RHR heat exchcnger component (such as the shell) impceting the edge of the relatively small hatch opening at one of these intermediate levels and tipping over onto the floor at that elevation.
The only safe shutdown equipment whose loss could potentially result in an incbility to cchieve and maintain safe shutdown that could potentially be impacted at e!evotions 326' or 300' is piping associated with the Emergency Service Wcter System cnd Reactor Building Closed Cooling Water System at the 300' elevation. Either of these systems con provide cooling to the RHR pumps.
Imocet of the piping ef interest covid result in an inab;lity to provide cooling to the pumps from either of these systems.
Scie shutdown equipment in the RHR beat exchanger cubicles at the 272' elevation whose failure could potentially result in on inability to cool the core include RHR service water piping. Loss of this piping could result in cn incbility to remove decay hect with the RHR heat exchcngers. As indicated above in the discussion for Region 7, the RHR heat exchangers are required to remove decay heat in the long term in all of several different possible system arrangements for accomplishing decay heat removal. Because the ability to accomplish long term removal of decay heat is threctened by potential lood drops in these regions, no heavy loads will be carried in these regions unless the plant is in the Refueling Condition, i.e. vessel head removed, reactor cavity filled, and spent fuel pool gate opened.
As indicated in the general discussion above regarding safe shutdown functions and systems in various plant conditions, there are a number of ways to successfully cool the core when the plant is in this condition. With these restrictions on load handling operations, the consequences of postulated lood drops in Regions 9E and 9W were determined to acceptably comply with NRC evaluation criteria.
40 j_ _
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References 1.
James A. Fitzpctrick Nuclear Power Plant Mcintenance Procedure No. 4.1,
" Disassembly of Recetor Vessel for Refueling", May 1980.
2.
" James A. Fitzpctrick Nuclecr Power Plant Maintenance Procedure No.
4.2, Recssembly of Reactor Vessel", July 1980.
3.
"Evoluction of Heavy Locd Hcndling Operations et Fitzpatrick Nuclecr Plant; Report No. I, Interim Actions and General Guidelines", TERA Corporction, September 1981.
4.
ANSI B30.9-1971, Slings, American Nctional Standard, Safety Stendcrds for Derricks, Heists, Hooks, Jceks end Slings, American Society of Mechanicci Engineers,1971.
S.
CMAA 70-1975, Soecifications for Electric Overbecd Traveling Crenes, Crone Mcnufacturers Association of America,1975.
6.
ANSI 30.2-1976, Overbecd and Gentry Cranes, American Nctiencl Stendcrcs, Safety Standcrds for Cableways, Crcnes, Derricks, Hoists, Hooks, Jacks and Slings, Americcn Society of Mechaniccl Engineers,1976.
7.
"lCES STRUDL, Structurci Design Language", McDonnell Dougics Automction Company, April 1981.
8.
Building Code Recuirements for Reinforced Concrete, ACI 318-77, American Concrete Institute, December 1977.
9.
Civil Engineering and Nuclear Power, Report of the ASCE Committee on impactive and impulsive Loads, Vol. V, American Society of Civil Engineers, September 1980.
i I
i
10.
Structural Analysis and Design of Nuclecr Plan? Facili9ies, Americen Society of Civil Engineers,1980.
11.
ACI 340-76, Code Reavirements for Nuclear Scfety-Reicted Concrete Structures, Appendix C "Special Provisions for Impulse and impactive Effects", American Concrete Institute,1976.
12.
Kennedy, R. P., "A Review of Procedures for the Anclysis and Design of Cortrete Structures to Resist Missile Impcet Effects", Journal of Nuclect Engineering and Design, Vol. 37, No. 2, May 1976.
13.
Mottock, A.
H., " Rotational Ccpecity of Hinging Region in Reinforced Concrete Secms", Flexurcl Mechanics of Reinforced Concrete, ASCE 1965-50 (ACI SP-12), American Society of Civil Engineers,1965.
14.
Corley, W.
C., " Rotational Ccpacity of Reinforced Concrete Becms",
Journcl of Structural Division, ASCE, Vol. 92, No. STS, Proc. Paper 4 39, Oct.1976, pp.121-146.
15.
" Design of Structures for Missile Impcet", Topical Report BC-TOP-9A, Bechtel Power Corporation, September 1974.
16.
Structures to Resist the Effects of Accidental Explosions, TM5-1300, Department of the Army, Washington, D.C., July 1965.
17.
Neville, A. M., Properties of Concrete, J. Wiley & Sons, New York,1975.
18.
Design of Structures to Resist the Effects of Atomic Weapons - Strength of Materials and Structural Elements, TM5-856-2, Department of the Army, Washington, D. C., August 1965.
19.
Personal communication between Professor William J. Hall and Howard A.
Levin, October 5,1981.
2
.y c
~ -
20.
Wcng, C. K. and Salmon, C.
G.,
Reinforced Concrete Desico, Intext Educational Publishers, New York,1973.
21.
Ferguson, P.
M., Reinforced Concrete Fundamentals, J. Wiley, New York,1973.
- 22.. Untrover, R. E. and C. P. Siess, " Strength and Behavior in Flexure of Deep Reinforced Concrete Beems Under Static and Dynamic Loading," Civil Engineering Studies Structural Research Series Report No. 230, University of Illinois, Urbanc, October 196l.
23.
Austin, W. J., et cl, "An investigation of the Behavior of Deep Members of Reinforced Concrete and Steel," Civil Engineering Studies Structural Research Series No.197, University of Illinois, Urbana, Jcnuary 1960.
24.
de Pcive, H.A.R., cnd C. P. Siess, " Strength and Behavior in Shear of Deep Reinforced Concrete Beams Under Static cnd Dyncmic Loading," Civil Engineering Studies Structurcl Research Series Report No. 231, University of Illinois, Urbcnc, Oct.1961.
25.
de Paiva, H.A.R., and W. J. Austin, " Behavior and Design of Deep Structurcl Members - Pcrt 3 - Tests of Reinforced Concrete Deep Beams," Civil Engineering Studies Structural Research Series No.1974, University of Illinois, Urcnc, March 1960.
26.
Winemiller, J. R. and W. J. Austin, " Behavior and Design of Deep Structural Members - Port 2 - Tests of Reinforced Concrete Deep Members with Web and Compression Reinforcement," Civil Engineering Studies Structural Research Series Report No.193, University of Illinois, Urbana, August 1960.
27.
Newmark, N. M. and J. D. Haitiwanger, " Air Force Design Manual -
Principles and Proctices for Design of Hardened Structures," AFSWC-TDR-62-138, December 1962.
3
o 28.
Crowford, R. E., et al, "The Air Force Monval for Desig, and Analysis of Hordened Structures," AFWL-TR-74-102, October 1974.
29.
Timoshenko, S. P., and Goodier, J. N., Theory of Elasticitv, McGraw-Hill Book Company, New York,1970.
30.
Bcker, Kovalevsky, and Risk, Structural Analysis of Shells, McCrow Hill, New York,1972.
31.
Reissner, "On the Theory of Thin Elcstic Shells",1949.
32.
Rocrk cnd Young, Formulations for Stress and Strain, McCrow-Hill, New York,1975.
33.
NUREG-06l2, Control of Heavy Loads et Nuclear Power Picnts, USNRC, July 1980.
34.
Effects of Imocet and Explosion, Summary Technical Report of Division 2, National Defense Resecrch Committee, Vol.1, Washington, D.C.,1946.
35.
Vassallo, F. A., Missile Impact Testinc cf Reinforced Concrete Penels. HC-5609-D-1, Calspen Corporation, Jcnvery 1975.
36.
Stephenson, A. E., " Full Scale TorncJo Missile impact Tests," Electric Power Research Institute, Final Report NP-440, July 1977.
37.
Beth, R. A. and Stipe, J. G., " Penetration and Explosion Tests on Concrete Slabs", CPPAB Interim Report No. 20, January 1943.
38.
- Beth, R.
A.,
" Concrete Penetration" OSRD-4856, National Defense Research Committee Report A-319, March 1945.
4 s
APPEtolX A This appendix contains Tcble 3 frem the Power Asthority letter to NRC d:'e:
October 15,1981 (JPN-St-32). Table 3 identif;es Reactor Building Crane hecvy loads.
In cddition to those loads listed in Tcble 3, postulcted drops of the Recirculctic Pump motor were c!so evolucted.
A-l
TAI)LE 3 I
REACTOR IlUILDING CRAtt i EAVY LOADS APPilOX.
APPLICABLE SAFETY WEIGHI LIF T LirTING INTERLOCK HANDLING j
LOAD CLASS (T ONS)
PitOCI OUf(ES EQUIPMEN T MODE Iti:STitlC TlONS 1.
Reoctor Vessel Head 1/3B 73 MP4.l A4.2 liend Strongback I'k>r mal Corry to minirnum
& Strongbock (7)
Turnbuckles height necessory
& Shockles obove vessel and refueling floor 2.
Drywell Head I/3B 48 MP4. l &4.2 Head Stronghock Normal Carry to minimum
& Strongbock (7)
Turnixickles height necessory A Shockles above vessel and refueling floor 3.
Steam Dryer &
I/3B 39 MP.4.1 A4.2 Dryer / Separator Normal Corry to minimum Sling Assembly (7)
Lif fing Sling height necessory above vessel 4.
Shroud Heod/Seporotor 1/30 43.S MP4.1 A4.2 Dryer / Separator Normal Carry to minimum
& Sling Assembly (7)
Lif ting Sling height necessary above vessel 5.
Reactor Cavity Shield 30 110 eo.
MP4.l &4.2 Slings, Turn-Normal Corry to minimum Plugs (5)& Shockles (7) lxickles A height necessary Siwickles above refueling floor 6.
Internals Storage Area 3B 40-50 ca.
MP4.l A4.2 Slings A Nor mal Corry to minimum Shield Plugs (3)
(7)
Siwickles height necessory obove refueling floor 7.
flefueling Slot 3A S.S co.
MPla.l A4.2 Slings A l'k>r m ol Corry to minimum Plugs (1)
(/)
Shorkles height neressor y about reIocling Iloor
s TAI)LE 3 (contimel)
APPROX.
APPLICAllLE SAFET3 WEIGi ll LIF T LIFTING INTEftLOCK HANDLING LOAD CLASS (TONS)
PROCEDURES EQUIPMENT MODE RESTRICTIONS 8.
Reoctor Vessel Head 1/3A 10 MP4.l A4.2 Reactor Head Normal Carry to minimum Thermal Insulation (7)
Insula tion height necessory Lif ting Rig above vessel and refueling floor.
9.
Reactor Vessel Head I/3 A 6
MP4. I A4.2 Reactor Head Stud Normal Corry to minimum Tensioners & Rig (7)
Tensioner Itig height necessary above vessel and refueling floor.
10.
Spent Fuel Pool 2
1.3 MP4. l & 4.2 Sling A Chain N/A Lif t with Chainfall.
Gates (2)
(7)
Fall 11.
Portable Rodiation 2/3A 14 MP4.1 &4.2 Slings A Shockles Normal Do not carry over Shield (Cattie Chute)
(7) reactor vessel.
Corry to minimum height necessary above refueling f loor.
12; Vessel Service I/3A 7
MP4.1 A4.2 Service Plot form Normal Carry to minimum Plot form (7)
Slings height necessory above vessel and j
refueling floor.
1 13.
Cleon Up Filter 2/3A 4.35 eo (7)
Slings A Shockles Normal Do not carry over Demineralizer reactor vessel.
Hatch Covers (2)
Carry to minimum height necessory above refueling f loor.
TABLE 3 (continued)
APPROX.
APPLICABLE SAFETj WElGHT LIF T LIFTING INTERLOCK HANDLING LOAD CLASS (TONS)
PROCEDURES EQUIPMENT MODE RESTRICTIONS 14.
Skimmer Surge Tonk 2/3A 3.7 en (7)
Slings A Shockles Normal Do not carry over Tonk Hotch Covers (2) reactor vessel.
Carry to minimum height necessory above refueling floor.
15.
RHR Heat Exchanger 2/3A 4.15 en (7)
Slings A Shackles l'k>rmol Do not carry over Hotch Covers (2) reactor vessel.
l Carry to minimum height necessory above refueling floor.
16.
New Fuel Storage 2/3A 3.75 (7)
Slings & Shockles t'k>r mal Do not carry over Voult Hotch Covers (3) reactor vessel.
Corry Io minirnum height necessory above refueling floor.
17.
Equipment Hatch 2/3A 0.5 eo (7)
Slings & Shockies Normal Do not carry over NW quadrant) reactor vessel.
Fkitch Covers (3)
Carry to minimum height necessary above refueling i
floor.
18.
Equipment Hotch 2/3B 1.3 en (7)
Slings & Shockles t'k>rmal Do not carry over SE quodront) reactor vessel.
Hatch Covers (5)
Carry to eninimum height necessory utmve refue ling iloor.
e.
TAllLE 3 (continued)
APPROX.
APPLICAllLE SAFET3 WEIGH T LIFT LIF TING INIERLOCK HANDLING LOAD CLASS (IONS)
PROCEDURES EQUIPMENT MODE RESTRICllONS 3
19.
Reoctor Building Crone 2/3B 3.1 (7)
N/A Normal Do not carry over Lood Block & Hook (when reactor vessel.
moving Do rwit carry over uni (xufed) equipment hatches except to make o lif t through hatch.
if over SE hatch, see footnote 3.
20.
Head Stud Rock 2/3A 1.5 MP4. l &I.2 Slings & Shockles Normal Do not carry over 4
(7) the reactor vessel.
Corry to minimum height necessory olx)ve refueling f loor.
3 21.
Shipping Cask 2/3B 3ts (6) (7)
Lif ting Yoke Normal See frocedure for CNS Is I 5 Supplied by Chem Cask Lif f '
4 f
Chem Nuclear i fondling,
3 22.
CNS4I:5 Cosk Liner 2/3A/3B I4 (7)
Sling provided Normal Carry to minimum with liner Cask height necessary g
I huulling,
oix)ve refueling floor. If over SE hatch, see footnote 3. Do not carry over reactor vessel.
4 1
TAHLE 3 (contintel)
OTHER APPROX.
APPLICABLE SAFETj WEIGHT LIFT LIFTING INTEftLOCK HANDLING LOAD CLASS (TONS)
PROCEDURES EOUlPMENT MODE RESiltlCllONS 3
23.
Spent Fuel Shipping 2/3B 70-110 (6) (7)
Lif ting Yoke Normal See Progedure for Cosk (Non-selected Cosk for LifI 4
as yet)
Handling 3
24.
Fuel Channel Crote 2/30 1.2 (7)
Mesh Slings fhrmal Corry to minimum height necessary above refueling floor.
If over SE Intch, see footnote 3.
Do not carry over reactor vessel.
3 l
25.
New Fuel Container 2/30 1.9 (7)
Slings fhrmal Corry to minimum l
height necessary almve refueling floor.
i Ifover SE hatch, see footnote 3. Do not carry over reactor vessel.
S 26.
RHR Heat Exchonger 2/3A 7.5 (7)
Slings & Shockles Normal Do not carry over Shell reactor vessel.
Corry 1o minimum height necessory above refueling floor.
1
TAI 3t 1
(continued)
A APPROX.
APPLICAllLE SAFETY WEIGH T Lir I LIFTING INil:RLOCK HANDLING j
LOAD CLASS (T ONS)
PROCI:DURES EQUIPMENT MODE RESTRICTIONS 27.
RHR Heat Exchanger 2/3A 20.5 (7)
Slings A Shockles Nor mol Do not carry over reactor vessel.
Carry to minimum height necessary above refueling floor.
28.
Hydrolaser 2/3A 2
(7)
Slings A Shockles t'k>r mal Do not corry over reactor vessel.
Corry to minimum height necessary above refueling floor.
29.
Recirculation Pump 2/3A 20 (7)
Slings A Shockles t'k>r mal Do not carry over Motor reactor vessel.
Carry to minimum height necessory above refueling f loor.
I NUREG 0612 defines a heavy load as one that weighs more than the combined weight of a single spent fuel assembly and its ossociated handling tool. For reference, the weight of a fuel ossembly, its associated handling tool, and chonnel at Fitzpatrick is opproximately 750 lbs.
2 Sofety Classes are defined in the response to item 3.o.
l 3
These foods are classified as 30 because of their potential for damaging equipment below the track boy floor at the 272' elevation. These loods must be lif ted over or up through the Reactor lloilding Equipment i kitch from the 272' elevation to the 369' elevation. The CNS 4-45 cask and Spent Fuel Shipping casks are also classified 311 hecouse of their potential for domoging equipment below the refueling and spent fuei pool floors, if dropped.
t 4
Interlocks restrict movement of the cask to the cask looding area when the cosk is over the spent fuel pool. See response to item 3.o.
S A RHR Heat Exchanger Shell or Tube flundle, if pulled for maintenance or replacement, must he raised to the 30' elevation from the 272' elevation through the ill III IlX i kitches.
6 Cask lifis will be governed by special lif t procedures that will be prepared in <wivonce of making the lifis.
7 All lifis will be addressed in a procedure governinq food Imndling operations by the Reactor fluilding Crone.
the dryer drop energy was determined to be 1.2 x 106 f t-lbs) and the dif ferent impacted items, both the dryer drop and the steam separator drop were evaluated.
in evcluoting the consequences of the steam dryer drop, it wcs necessary to determine the sequence of "f ailure modes" for the drop. This was determined based on locations of first failure in the impact system. That is, the weckest links were assumed to fail first and the energy absorbed cnd dissipated in that failure wcs calculcted and compcred to the total energy of the drop.
Similarly, the work done in other parts of the supporting structure in absorbing i
the impact of the dryer drop wcs calculated. In this manner, if the supporting structure could be shown to absorb and dissipate the energy of the dryer drop without dcmaging the fuel, the consecuences of the drop were cssumed to be cceeptable.
The analysis for the dryer drop indicates thct in actuality the weakest link in this drop is the crushing of the dryer essemblies themselves. This crushing of the dryer dissipctes the total energy of the dryer drop and thus no other dcmage to the supporting structure is predicted. In addition, the loads of this drop onto the supporting structure were evaluated and found to be acceptchle.
Although crushing of the dryer is predicted, in order to be consistent with the guidelines of NUREG-0612, on evaluation was made to determine whether the supporting system could withstand the energies of the dryer drop assuming that no crushing of the dryer took place. Again in this case, the supporting system, including the i
shroud and shroud support, were found to acceptcbly withstand the energy of the dryer drop.
In addition to evaluation of drop energies, a dynamic impact factor was calculated for both the dryer drop and steam separator drop. The methodology for calculating this impact factor was similar to that described above for the reactor vessel head drop. A dynamic impact factor was calculated for the dryer drop, and the resulting total dynamic load on impact was determined. When the dryer drops, it lands on the four dryer support lugs. The calculated stress in these lugs exceeds the allowable shear stress for the lug materici. Therefore, the four dryer support lugs are assumed to fail. Failure of these four s'upport lugs results in the dryer dropping onto the steam separator. As was mentioned above, the energies of this drop were evolucted and found to be occeptoble. In w
14Q
addition, the stresses through the load pcth of the supporting system (i.e., steam separctor, shroud, shroud support, recctor vessel, and skirt) were evolucted cnd found to be acceptable.
Based on the results of the recctor vessel head drop and steam dryer drop scencrios enclyzed to date, both from the standpoint of stresses and energies, reactor vessel integrity is predicted and no fuel dcmage is expected. However, additional RPV head drop scencrios will be evolucted to provide additionc!
assurance that vessel integrity con be maintained.
An onelysis for a postulated drop of the steam separator assembly was also performed, in a menner similcr to that for the steam dryer drop. However, wherecs the consequences of the steam dryer drop were determined to comply with the criteric of NUREG-0612, the results of our onclyses performed to date indicate thct the consequences of the postulated steem separator drop could have a potential for cousing fuel domcge. Since the use of odditional safety margins is desirable, we will undertake additionc! investigations to ottempt to demonstrate that the likelihood of a steam separator drop, os onelyzed, is sufficiently small or to suitcbly constrain the pcrameters of the drop to allow acceptable consequences to be demonstrated, in addition, to assure that fuel integrity is maintained for other types of drops, investigations of the consequences of a postulated drop of the portable radiction shielo or the refueling slot plugs into the recctor vessel were undertaken. In both cases, the consequences were determined to be conservatively bounded by the results of the steem dryer drop analysis. Therefore, reactor vessel integrity was predicted and no fuel damage is expected.
15 i
1 e
2.3 SPECIFIC REQUIREMENTS FOR OVERHLAD HANDLING SYSTEMS OPERATING IN PLANT AREAS CONTAINING EGUIPMENT REQUIRED FOR REACTOR SHUTDOWN, DECAY HEAT REMOVAL, OR SPENT FUEL POOL COOLING j
NUREG-0612, Section 5.l.5, provides guidelines concerning the design and operction of load-handling systems in the vicinity of equipinent or components required for scfe reactor shutdown and decay heat rer novol.
Information I
provided in response to this section should be suf ficient to demonstrate that i
adequate mecsures have been token to ensure that in these creas, either the likelihood of c drop which might prevent safe reactor shutdown or prohibit continue decay heat removal is extre.nely small, or that damage to such equipment from loads will be limited in order not to result in the loss of these safety-related functions. Cranes which must be evoluoted in this section have been previevsly identified in your response to 2.l-1, and their londs in ycur response to 2.1-3-c.
ITEM 2.3-1 Identify any cranes listed in
.1-1, cbove, which you have evaluated as having suf ficien, design features to make the likelihood of a load drop extremely smcIl for all Icods to be ccrried cnd the basis for this evaluation (i.e., complete compli-cnce with NUREG 0612, Section 5.1.6. or partici compliance supplemented by suitchie c!ter nat ivt nr adJitional design features).
For each crcne so evoluoted, provice the food-hcr. ding-system (i.e.,
crone-lood-combination) information 4
specified in Attachment 1.
RESPONSE
The Reactor Building crane was evoluoted to industry standards CMAA 70-1975 (reference 5) and ANSI B30.2-1976 (reference 6). It was found to meet those standards, with two exceptions which were justified in reference 3.
Notwithstanding the fact that the lif ting system including the Reactor Building crone, lif ting slings and strongbacks, complies with the intent of applicable industry standards and possesses demonstrated margins to failure, rather than relying on the reliability of the lif ting system, on evoluotion has been performed to ossess the consequences of postulated drops of heavy-loods.
Therefore, although these heavy load drops need not be postulated, even if they were to occur, their consequences have been evoluoted.
16
~,_
The only cose where load handling reliability was considered was with respect to the main hoist lood block and hook. NURFG-0612 (reference 3'4) requires that the food block cad hook be considered as a heavy foad. The load block is used for handling numerous loads, including the reactor vessel head, drywell hecd, shield
. plugs, and the dryer and sepcrator units. In moving these foods, the hook, load block, rope, drum, sheave cssembly, motor shaf ts, gecrs, and other load becring members cre subjected to significent stresses approaching the food rating of the crcne. By comparison, these components are subjected to a considerably smc!!er locd when only the hook and load block cre being moved. Based on this,it is not considered feasible to postulate a random mechanical failure of the crane load becring components when moving the crene load block alone.
The only feasible failure modes for droppino of the main book an.1 Inad block would be:
i l}
A control system or operator error resulting in hoisting of the block to a "two blocking" position with continuad hoisting by the motor and subsequent pcrting of the rope (this situction con be prevented by operator action prior to "two blocking" or by cn upper limit switch to terminate hoisting prior to "two blocking"); cod 2)
Uncontrolled lowering of the load block due to tailure of the holding brake to function (the likelihood of this can be made smcil by use of redundant holding brakes).
The Fitzpatrick Reccior Building crane is provided with two diverse upper limit switches to interrupt power to the hoist motor prior to "two blocking." When power is removed, holding brckes are automatically applied. One of the two limit switches is a geared limit switch driven off the drum shaf t. The other is a counter weight switch that is released when the load block comes up ogainst a trip bar; the trip bar will stop power to the hoist below the low po.nt of the sheave assembly.
The holding brckes are solenoid released, and spring applied on loss of power to the solenoid. Two holding brakes are provided, either of which has sufficient copocity to hold the rated load (each broke is 150% of full motor torque).
\\
Additionally, inspection and maintenance procedures assure that the limit j
l switches and holding brakes are functional and properly adjusted.
I7
With the provisions described above, the two diverse litnit switches will reduce the likelihood for "two blocking" and the two hohling brakes will reduce the likelihood of uncontrolled lowering of the loud blo:k. Based <>n these features, it is concluded that a drop of the load b!o-4 and hook is of suf ficiently Ic.v likelihood that it does not require load dra;' coolyses.
Nonetheless, en enclysis of a load block und hook drop from the highest possible ccrry height onto the operating /ref ueling floor wcs perfor:ned to verif y the copcbility of the floor to withstand the impcci of such a drop. The resslts of the onclysis indiccie that while concrete secbbing on the underside of the floor is predicted, no gross failure or penetration will result.
The consequences of secbbing have been considered in the systems evoluotions and were found to be cecepf cble.
Therefore, cithough drop of the load block and hook need not be postulated, even if they were to drop on the operating / refueling floor, the consequences cre ceceptcble.
18
8 ITEM 2.3-2 For any cranes identified in 2.!-l not designated as single-failure-proof in 2.3-1, o comprehensive hczord evolvotion should be provided which includes the following information:
The presentation in a metrix formet of all heavy loads and a.
potentici impact creas where damage might occur to safety-related equipment.
Heavy loods identification should include designction and weight or cross-reference to information provided in 2.1-3-c. Impact creas should be identified by construction zones and elevations or by some other methods such that the impact crea ccn be located on the plant general crrcngement drawings.
Figure i provides o typical matrix.
b.
For each interaction identified, indicate which of the load and impact crea combinations con be eliminated because of sepcration and redundancy of safety-related equip-ment, mechcnical stops end/or electrical interlocks, or other site-specific considerations.
Elimination on the basis of the aforementioned consideration should be supplemented by the following specific information:
(1)
For load /torget combinations eliminated because of separation and redundancy of safety-related equip-ment, discuss the bcsis for determining that loads drops will not affect continued system operation (i.e., the cbility of the system to perform its safety-related function).
(2)
Where mechanicci stops or electrical interlocks are to be provided, present details showing the crecs where crcne travel will be prohibited. Additionally, provide o discussion concerning the procedures that cre to be used for authorizing the bycssing of interlocks or removable stops, for verifying that interlocks are functional prior to crane use, and for verifying that interlocks are restored to operability after operations which require bypassing have been completed.
(3)
Where load /torget 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 validity of such considero-tions.
C c.
For interactions not eliminated by the analysis of 2.3-2-b obove, identify any handling systems for specific loods which you have evaluated as having sufficient design features to make the likelihood of a lood drop extremely small and the basis for this evaluation (i.e., complete 19
c:mpliance with NUREG 0612, S ction 5.l.6, or partial complicnce supplemented by suitable citernative or addi-tional design features). For each so evaluated, provide the load-handling-system (i.e., crone-lood-combination) information specified in Attachment 1.
d.
For interactions not eliminated in 2.3-2-b or 2.-3-2-c, above, demonstrate using appropriate analysis that demoge would not preclude operation of sufficient equip-ment to allow the system to perform its safety function following a lood drop (NUREG 0612, Section 5.1, Criterion IV). For each analysis so conducted, the follow-ing information should be provided:
(1)
An indication of whether or not, for the specific load being investigated, the overhead crane-handling system is designed and constructed such that the hoisting system will retain its load in the event of seismic accelerations equivalent to those of a safe shutdown earthquake (SSE).
(2)
The bcsis for cny exceptions taken to the analytical guidelines of NUREG 0612, Appendix A.
(3)
The information requested in Attachment 4.
RESPONSE
The reactor building crane is normally used for maintenance operations which inc!vde moving of items above the operating / refueling floor at 369'-6" elevation, and movement of equipment from the track floor at 272' elevation up the SE equipment hatch to the operating / refueling floor.
In addition, the reactor building crane con be used in a " cask handling mode" as discussed in our initic! submittal responding to the NRC letter of December 22, 1980 (letter from George T. Berry to Darrell G. Eisenhut dated October 15, 1981).
Accordingly, evaluations of heavy lood handling operations at Fitzpotrick require considerations of drops onto the operating / refueling floor at 369'-6" elevation, drops into the reactor vessel and spent fuel pool, drops into the mternals storage pit, and drops onto the 272' elevation floor at the equipment batches. Evalua-tions for drops into the reactor vessel and spent fuel pool were discussed previously in response to item 2.2.4.
The evaluation of heavy lood handling operations at Fitzpatrick was performed by reviewing those heavy loods which could be corried over each region of the 20 l
r 1
r: actor building.
The rgoctor building was subdivided into nine regions of interest, covering the crecs where heavy loads could be dropped (Figures 2-10).
Accordingly, our responses to the above requests for information cre provided below on a region by region basis.
A combination of systems evoluctions and structural enclyses were utilized to verify that damage following a postulated load drop would not preclude operction of sufficient equipment necesscry to perform scfe shutdown (NUREG-0612, Section 5.1, Criterion IV). A drop of each heavy load carried by the reactor building crone was postulated to occur onto the operating /ref;eling floor at elevation 369'-6" or equipment hatches, as cppropriate.
Worst ccse drop scencrios were evolucted, so that the consequences of a postulated drop of cny of the heavy loads handled (see Appendix A) cre bounded by the results presented herein.
Systems Evoluotion Methodology - Safe Shutdown Evaluation As part of the evaluation of heavy load handling operations at Fitzpatrick, a
number of potentici load drop regions in the reactor building were addressed by performing systems evoluctions. The objective of the systems evaluations 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 possible load drop. The results of the systems evaluations are summcrized below in the discussions for each region.
In order to demonstrate the ability to safely shutdown and cool the core, it was necessary to (1) identify the safety functions required to achieve safe shutdown, (2) identify the plant systems required to occomplish these functions,(3) identify the equipment that could potentially be lost if a food drop were to occur in certain plant areas (designated as Regions), and (4) determine the resultent effects of the loss of this equipment on the safety functions required to achieve safe shutdown.
21 4
m.
w r
--- - - - ~
l Plcnt Conditions f
To determine the functions that must be accomplished to achieve and maintain sofe shutdown,it was assumed for most regions investigated that the reactor was 3
of 100% power at the time the load drop was postulated. For 'tegions 9E and 9W, i
however, it was assumed that movement of the load of int: :est in these regions would only be performed with the reactor in the shutdo.vn and cold condition.
Accordingly, the plant conditions associated with the cold shotdown ond/or i
Refueling Conditionl/ were assume d as initio! conditions for evaluating drop l
t consequences in these regions. The functions required to be accomplished and systems included in flie embutica to accomplish these f onctio is are desribed 1
below.
Saf e Shutdown Functions and Svste:ns In order to cecomplish scfe shatdown from 100% power, the following functions must be performed:
Reccior Scram or Shutdown Monitoring of Critical Plant Perometers I
Core Cooling (Initial)
Depressurization/Mokeup Extended Cooling 4
All of the cbove functions (except Scram and Extended Cooling) con be accomplished as port of normal plant cooldown with non-safety systen s such as the feedwater, condensate, and circulating water systems.
Nonetheless, no credit was taken for non-sofety systems in performing the systems evoluotions, i.e., the ability to accomplish safe shutdown and core cooling was evaluated assuming the use of safety systems only, except for those cases where the Refueling Condition was relied upon.
If The Refueling Condition referred to here is defined for purposes of the systems evaluations to be reactor head removed, reactor cavity (and possibly the Storage Pit) filled and Spent Fuel Pool Gate open.
4 9
22 r
The functiors and sp:cific systcms that could b2 relied on to cchieve end mcintain safe shutdown are indiccted in Figure 11.
if the plant proceeds to the Refueling Condition,l/ then several cooling modes are possible. Exemples cre: (!) RHR Shutdown Cooling, (2) Spent Fuel Pool Cooling and Ciecnup System Cooling and (3) RHR Fuel Pool Cooling. Certain of these cooling modes rely on non-sofety equipment. However, they are consi-dered since in the Refueling Condition significent time is availcble to estcblish alternate cooling. In addition, if cny or all of the cooling systems referred to above were lost, the core and spent fuel in the storage pool would continue to be cooled by the body of water in the pool and the reactor cavity. All that is necesscry is to provide a source of mckeup water to replenish cny loss of inventory. Mckeup could be provided by hoses or by other crrcngements from any of a number of different available water sources, if required. Accordingly, if a load is only handled during this plant condition, there is no single load drop scencrio that could result in inability to cool the core.
Steos in the Systems Accrocch The following summcrizes the steps thct were performed in the systems evaluations for each function / system required for safe shutdown:
1)
Identify the system (including any support systems) components of interest.
2)
For each potential load impoet region evaluate potential for damage / loss of system components.
If equipment could be impacted, assume it is lost.
3)
Compare system equipment required (Item I), with equip-ment lost (Item 2), and determine if the function for which the system is relied on could be lost.
4)
Review for other potential system interactions based on equipment demoged/ lost and determine if function could be lost.
5)
If the system evoluotion reveals that the system could occomplish its safety function following a lood drop into the region of interest, then no further evaluation is necessory.
23
6)
If th2 systsm Gvoluction revcals that tha system function could potentially be lost, then evolucte the possibility of relying on citernctive safety systems to accomplish the some function following a postulated load drop into the region.
7)
The overo!! safe shutdown conclusion regcrding a parti-culcr region is the composite for that region of the conclusions for all the systems required to accomplish the safe shutdown functions.
Structural Evaluation Methodology Each of the heavy loads corried by the reactor building crone have been evolucted to identify loads which control local response (e.g. penetration, secbbing, spelling, perforation, etc.); loads that control overall structurci response (e.g. Icrge inelastic deformations or obrupt failures of principal structural members, etc.); and/or loads that may induce behavior that exhibits combined response such that either overall or local failure modes would control.
The results of this evoluction are tabulated in Table 2.
In each region where local response was evolucted the load drops were onelyzed to verify that sicb perforation (i.e. penetration entirely through the floor sicb) did not occur.
Seebbing of the concrete deck backface wcs evaluated for cl! loads. In cases where postulated drops were predicted to produce this effect, equipment and systems below that crec which could be impacted by the scabbing were assumed to be lost for the purpose of performing the systems evoluotions.
Where the controlling modes of the heavy load drop response were determined to be "overall structure" response modes, these load drops were evolucted to verify that gross and intolerable distortions of the primary structural members did not occur. By verifying that gross and possibly propogating failures would not occur for these food drops, the consequences of the food drops could be shown to be limited to scabbing.
24
TABLE 2
SUMMARY
OF CONTROLLING STRUCTURAL BEHAVIOR RESULTING FROM POSTULATED REACTOR BUILDING CRANE HEAVY LOAD DROPS CONTROLLil1C MODE OF
RESPONSE
APPROX.
WElGHT OVERALL LOAD (TONS)
STRUCTURAL LOCAL 1.
Reactor Vessel Head 73 X
& Strongback 2.
Drywell Head &
48 X
Strongback 3.
Steam Dryer & Sling 39 X
Assembly 4.
Shroud Head / Separator 43.5 X
'& Sling Assembly 5.
Reactor Cavity Shield 110 ec.
X
. Plugs (5) & Shackles 6.
Internals Storage Area 40-50 ea.
X Shield Plugs (3) 7.
Refueling Slot 5.5 ea.
X Plugs (3) 8.
Reactor Vessel Head 10 X.
Thermal Insulation 9.
Reactor Vessel Head 6
X Tensioners & Rig 10.
Spent Fuel Pool 1.3 X
Gates (2)
II.
Portable Radiction 14 X
X Shield (Cattle Chute)
- 12. ~ Vessel Service 7
X Platform 25
TABLE 2 (continued)
CONTROLLING MODE OF
RESPONSE
APPROX.
WEIGHT OVERALL LOAD (TONS)
STRUCTURAL LOCAL 13.
Clean Up Filter 6.35 ec X
Demineralizer Hatch Cover s (2) 14.
Skimmer Surge Tonk 3.7 ec X
Tonk Hctch Covers (2) i 15.
RHR Heat Exchcnger 4.15 eo X
Hatch Covers (2) 16.
New Fuel Storage 3.75 X
Vault Hatch Covers (3) 17.
Equipment Hatch 0.5 ea X
NW quadrant) i Hotch Covers (3) 18.
Equipment Hatch I.3 eo X
SE quadrant)
Hatch Covers (5) 19.
Reactor Building Crane 3.1 X
Lood Block & Hook 20.
Head Stud Rock I.5 X
21.
Shipping Cask 34 X
X CNS 4-45 22.
CNS 4-45 Cask Liner 4
X 23.
Spent Fuel Shipping 70-110 X
X Cask (Non-selected as yet) 24.
Fuel Channel Crate 1.2 X
25.
New Fuel Container 0.5 X
26
TABLE 2
. (continued)
CONTROLLING MODE OF
RESPONSE
APPROX.
WElGHT OVERALL 3..
LOAD (TONS)
STRUCTURAL LOCAL
- 26. - RHR Heat Exchcnger 7.5.
X Shell 4
27.
RHR Hect Exchanger 20.5 X
X 28.
Hydroicser 2
X
]
29.
Recirculation Pump 20 X
X Motor i
4
.f.
l 27 1
.e
,=-*-=n,*vvery vi my 7--
' in the onclyses of possible sicb perforation, procedures recommended in references 9 and 10 were followed. The modified National Defense Resecrch Committee (NDRC) formula (reference 35) was chosen because it has been shown to give the best fit with available experimental data (references 36 and 37). The NDRC formula predicts the possible depth of penetration of a solid missile. In order to determine the thickness of the reinforced concrete needed to resist impact without perforation or scabbing the Army Corps of Engineers formula was used (reference 38) in conjunction with the NDRC penetration formula. In addition, a 10% margin on concrete floor thickness was conservatively coplied as i
recommended in reference 9.
Although limited penetration and secbbing were predicted for the set of bounding heavy locd drops considered, in no case was the elevation 369'-6" sicb predicted to be perforated for normal carry heights.
For those drops which were determined to be controlled by overall structural response, the methodology used to evolucte the consequences of the heavy load drops was one of characterizing structural behavior in terms of the availcble 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 inelcstic deflections without complete collapse, while the brittle mcde may result in partic! failure or total collapse. The available internal strain energy that can be obsorbed by the floor system without reaching those limits of unacceptable behavior is balanced against the externally applied energy resulting from the load drop. It was assumed ire those calculations, that momentum is conserved and the kinetic energy of the. drop drives the mass of the floor and induces strain. As on additional conservatism, no credit was taken for potential sources of energy dissipation such as concrete crushing and penetration.
A four-step iterative step wise linear static analysis was performed using the STRUDL computer code (reference 7) to determine force-deflection for impor-tant points in the structural model. The computational procedure of the analysis is based on a network interpretation of the governing equations, the principal feature of which is the segmentation in processing of the geometrical, mechani-col, and topological relationships of the structure. In this case, a plain grid 28
model was developed which allows loads and deflections normal to the plane of l
the grid, and rotations above the axes lying in the plane.
The basic steps in the STRUDL computation procedure were os follows:
1)
The stiffness matrix is determined for each member and finite eterrent. Members cre considered as contilivers.
2)
If any member releases cre specified, the local member stiffness mctrix (step l} is modified.
3)
The applied member cnd element loads, if any, are processed.
4)
When free joints or released support joints exist, the structural stiffness matrix is assembled in a global coor-dinate system.
5)
The load vector is ossembled cnd the global stiffness motrix in the load vector are modified to account for joint releases.
6)
When there are free joints or release support joints, governing joint equilibrium equctions are solved for the joint displacement.
7)
The induced member distortions, member end forces, element strains and stresses are computed by back-substitution.
The model is successively loaded until the moment capacity of any section is exceeded.
This moment capacity is defined by Chapter 10 of ACI 318-77 (reference 8). At this point the model is reconstructed, incorporating oppro-priate rotational member releases. Greater multiples of the drop load cre then applied to the new model until moment copccities are reached at other sections.
This procedure continues until the ultimate food of the sicb/ grid system is reached.
Generally, the ultirnate food of a sicb/ grid system is reached prior to exceeding the hinge rotational capacity of porticular sections, provided that on unstable mechanism has not formed. This was found to be the case in the analysis for Fitzpatrick heavy lood drops. The hinge rotational capacity was used is a criterion to set a maximum allowable level of deflection for the slab /g,.J 29
system. Th2 hinge rotational capacity for concrete structures was d2veloped in references 11 and 12 based on test results given in references 13 and 14.
Rotetions of the magnitude suggested in the cbove references and used for these l
analyses result in cracking which is confined to a region below (above) the tensile reinforcement.
Genero!!y speaking the section will remain intcet with no crushing, spelling or scabbing, due to flexure; however, scabbing may occur as a result of shock wave motion cssociated with the reflection of tensile waves from the rect surface or shect plug formation. Therefore, it has been conservatively cssumed that secbbing does occur due to these drops.
The load / deflection history up to the point of the maximum load coupled with the maximum allowcble deflection, defines the maximum level of strain energy adsorption, provided that a shear failure has not occurred.
At each load increment in the onelysis (specified in terms of load amplification factors) the sheer stress at limiting sections is checked and compared to allowcbles as specified in Chcpter 11 of ACI 318-77 (reference 8).
The load cmplification factors are utilized cs a convenient method of determining the sicb/ grid resistance and should be differentiated from dynamic load factors.
The moment diagram at the ultimate load and the slab / grid deflections os the ultimate load is reached, cre each compared against allowables. In the cose of the deflection, the allowcble deflection is limited by the rotational capacity of the member. Integrating the sicb/ grid force-deflections under the loaded region, and performing on energy balance, the allowable drop heights for each heavy load considered are determined.
In addition to the conservatisms previously mentioned, the following conserva-tisms are also inherent in the methodologies used in the structural evaluations:
1)
Static material strengths for concrete and steel were used, although test data shows that this property increases with the increased strain rates associaied with dynamic loadings.
For example, references 15 and 16 recommend dynamic increase factors of 1.25 for compres-sive strength of concrete and 1.20 for the flexural, tensile and compressive strength of structural steel.
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i 2)
Drsign (minimum) material properties for concrete and steel were used. No increase was taken for the acfng of l
concrete, which con amount to a factor of up to 1.35 l
(reference 17) of increased strength. Also, the average strength for structural steel is nearly a factor of 1.25 (reference 18) highered in the minimum yield requirement specified by ASTM. While these factors above minimum code strength exist and contribute to structural margins, they were not used in the evoluction.
3)
The criteria for hinge rotational capacity that was used corresponds to support rotations of the order of 2 degrees, with minimum cracking and no crushing or secbbing. To meet necesscry performance requirements (i.e. halting propcgating failure), larger rotations in the range of 5 to 12 degrees could be tolerated. Experimental observations i
(reference 19) suggest even further capability fer well designed and well cnchored sicbs.
Use of these larger rotational copcbilities would have resulted in greater ene%f capcbilities of the grid system.
4)
The enclysis used ACI 318-77 o!!oweble shear stresses. A significant body of data suggests the existence of higher sheer copcbilities on the order of 10 Vf'c to 20 Vf'c (references 20 through 28). It is expected that the sheer capcbilities of the beams at elevation 369' would tend to be in the higher end of the range since the majority of the becms cre " deep". Deep beams behave as tied orches with significant reserve capacity.
5)
The cnolysis neglected the two-way resistance capcbility of the slob. It is expected that the sicb would contribute increased strength, particularly at Icrger deformations.
6)
The food was distributed directly under the drop.
In reality a more favorable load distribution would exist due to the food distribution capcbility of the sicb.
7)
No credit was taken for local energy dissipation asso-ciated with any crushing of the foods or the immediate surface of the floor.
Regions I and 2 Region I (Reactor Vessel) and Region 2 (Spent Fuel Pool) evaluntions were discussed previously in response to item 2.2 4.
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1 Recion 3 Region 3 (Internals Storage Pit) was evoluoted for the bounding load of the drop of ne steam dryer.
A structural evaluation was performed to determine whether overall structural failure, or local damage, could impact equipment necessary to accomplish and maintain safe shutdown.
The results of the onelysis indicated that overall structural integrity of the storage pit would be maintained for a postulated drop of the steem dryer.
Furthermore, it wcs concluded that a gross brecch of the leak tight integrity of the pit would not occur, and that if in the unlikely event that leakoge should occur, it would be extremely minor and would be limited to insignificent dripping through small cracks.
Scobbing of the concrete under the pit was assumed to occur.
Therefore, demcge to equipment in this region (at elevation 326' below the storage pit) was evolucted on the bcsis of potential secbbing of concrete from the underside of the storage pit floor.
The evaluation determined that no safe shutdown equipment whose failure could result in on incbility to accomplish cnd meintain safe shutdown could be impacted in this region.
Therefore, based on the structural and systems evaluations, the consequences of heavy load drops into Region 3 were determined to be acceptable.
The steam dryer assembly is mounted in the reactor vessel above the steam separator assembly. The dryer is cylindrically shaped, weighs 39 tons, and is 316 inches long and 214 inches in diameter.
The separator is also cylindrically shaped, weighs 43.5 tons, is 201 inches long and 200 inches in diameter. Each unit must be lifted less than 6 feet above the storage pit floor when moving it into and out of the storage pit. The drop of the dryer was determined to be more controlling, because it free falls through air, versus the separator which falls through water.
Although each unit is highly crushoble, no credit was given to energy obsorbtion due to crushing. Instead, the units were conservatively assumed to be infinitely rigid.
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The storage pit slab is 20 feet by 40 feet in dimension, and 5 feet thick. A yield line onelysis assuming a uniformly loaded circular fan was conducted to determine the ultimate resistance of the stab. The deflection of the slab was calculated on the basis of slab rotational capacity. The absorbed energy was then obtained by integrating the deflection under the load.
The maximum allowable carry height was therefore conservatively determined to be more then twice the normal carry height of the units.
Reaion 4 Region 4 includes the entire south half crea of the 369'-6" elevation floor. Locds identified in Appendix A were evaluated for postulated drops over this area. On the basis of load weights, dimensions and drop heights; bounding loads were evaluated for both the overall structural mode and the local response mode of behavior previously discussed. For the overall structural mode, the bounding load drops were determined to be those for the reactor cavity shield plugs, and topple-over of the fuel cask. The five reactor cavity shield plugs (A, B, C, D, E) shown in, figure 12 each weigh approximately 110 tons. The plugs are 6'l" thick and range between 33 to 38 feet in length, and 7 to 10 feet in width. The plugs are carried from the reactor cavity cpproximatley 6 inches above the refueling floor to the laydown crea shown in figure 12.
The shield plug lifting slings were previously evaluated (reference 3) and found to meet industry standard ANSI B30.9-1971 (reference 4). In addition the reactor building crane was evaluated to industry standards CMAA 70-1975 (reference 5) and ANSI B30.2-1976 (reference 6), and found to meet those standards with two exceptions. Justification for those exceptions wcs provided in reference 3.
Notwithstanding the fact that the lifting system, including the crane and slings, complies with the intent of the applicob'e industry standards and possesses demonstrated margins to failure, on evaluation was performed for o postulated drop of any of the shield plugs onto the reactor building refueling floor at elevation 369'-6". The worst case drop was considered to be that of Plug C offer the other four plugs had already been moved to their laydown oreo positions.
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The concrete floor et elevation 369'-6" is generally 15 inches thick except at selected crecs where it is 24 inches thick and at the reactor head Icydown crea where it is 30 inches thick. The floor sicb is supported by a grid system of continuous concrete becms which very in depth from 40 to 60 inches and in width from 26 to 30 inches, with reinforcement ratios of 0.12 to 1.6 percent. The sicb/ grid is supported cround its periphery by sheer walls and at intermediate points by concrete columns.
The overall structural response methodology discussed previously was used to verify that the concrete floor possessed sufficient internal strain energy capabilities to withstand the drop of the shield plug.
It was assumed that momentum is conserved and kinetic energy of the drop drives the mass of the floor and induces strain. As en additional conservatism, no credit wcs taken for potential sources of energy dissipotion such as concrete crushing and penetro-tion.
The results of the bounding drop evaluation determined that the slab / grid system possessed sufficient energy obsorbing capacities to withstand the drop of the cavity shield plug from heights significently exceeding the normal carry heights.
In addition, the effects of local structural response from smaller and perhcps lighter drops were also evolucted in this region. Loads such cs the portable radiction shield, refueling slot plugs, various hatch covers, and fuel channel crates were evaluated to assess the acceptability of postulated drops os limited by the concrete deck capability to resist perforation. The analysis methodology was as described previously for local response evaluations.
For all such evaluations, in no cose was the elevation 369'-6" sicb predicted to be perforated for these loods.
Although spent fuel casks are not currently handled at Fitzpatrick, a drop of a cask was postulated to occur on the refueling floor at Region 4.
Since fuel assembly and low-level radioactive material shipping casks have not been selected to date, the 34 ton (maximum loaded weight) Chem-Nuclear Systems, Inc. C.N.S. 4-45 cask and on as yet unspecified 110 ton fuel assembly cask were evaluated. The 110 ton cask was assumed to measure 5 feet in diameter and 18 34
e fcot in isngth. The C.N.S. 4-45 cask is 1731/8 inches long and 421/2 inches in dicmeter, except for 315/8 inches at each end which is 401/2 inches in diameter.
Each end of the cask hcs a cover and on impact limiter.
However, for conservatism, the energy absorbing effects of the impact limiter wcs not accounted for.
The ecsk was cssumed to be moved over the 369'-6" elevation deck to and from the equipment hatch, spent fuel, and ecsk washdown (head storage) crea. The sicb is 24 inches thick along the travel path between the spent fuel pool and the equipment hatch. Three concrete becms span the approximately 30 feet between the hatch cnd the spent fuel pool, and are the principai lood ccrrying n' embers.
Both overall and local structural response modes were evaluated for a postulated drop from the normal carry height of 6 inches followed by c topple-over onto the deck. The structural methodologies were as described above. Perforation of the sich is not predicted for either cask. Also, the postulated vertical drop does not cause unacceptcble hinge rotation or sheer failure, os defined by the criteria described ecrlier.
However, for topple-over of the casks, a prelimincry conservative analysis predicts the rotational capacity criteria to be exceeded.
Therefore, overall structural failure con not be precluded.
While these initial evoluctions have indicated that postulated cask drops onto this region of the refueling floor could result in overall structural failures, several possible options for preventing this result have been identified and cre being evaluated. The results of these evaluations will be submitted to the NRC when they have been completed. Accordingly, no cask handling in this cree will be performed until on acceptable solution hos been implemented.
Since all of the structural evaluations, except those for cask topple-over, indicated that damage con be limited to concrete scabbing onto the elevation immediately below the refueling floor, systems evaluations were performed to determine the effects on safe shutdown of damage to equipment at elevation 344' in this region. Loss of the only safe shutdown equipment that could be impacted in this region could result in the inability to utilize one of the two redundant Low Pressure Coolant injection (LPCI) loops to provide makeup during 35
rcactor depressurization/ cooling. The other loop of LPCI would be unoffected, in addition other systems, such as Low Pressure Core Spray could be utilized to accomplish this function. As indicated cbove, large casks will not be handled over this crea of the refueling floor until on acceptable s,olution hos been implemented.
Therefore, the consequences of postulated load drops in Region 4 were evaluated and, except for the ecse of cask topple-over, were found to acceptably comply with NRC evoluotion criteric.
Recion 5 Region 5 is the r.ortheast quadrant of the refueling floor at elevation 369'-6".
Structural evoluotions of this region were determined to be bounded by those evaluations performed for Region 4, since the floor sicb of this region is 24 inches thick versus 15 inches thick as evaluated for Region 4, and the heavy load carried in Region 5 cre similar and bounded by those evaluated for Region 4.
Therefore, structural domcge in Region 5 is limited to scabbing of concrete under the floor sicb. The effects of damage to equipment at elevation 344' below this region was evaluated on the basis of potential for this secbbing. No safe shutdown equipment whose failure could result in on inability to accomplish and maintain safe shutdown could be impacted in this region. Therefore, the consequences of load drops in region 5 were determined to acceptably comply with NRC evaluation criteria.
Region 6 Region 6 includes the northwest quadrant of the refueling floor, except for the equipment hatch and the internal storage pit areas. The heavy loads corried in this area were determined to be similar to those carried in Region 4 and therefore the structural evaluations previously described for Region 4 apply to Region 6 also.
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