NUREG 0612CONTROL OF HEAVY LOADS

R. E. GINNA NUCLEAR POWER PLANT

ROCHESTER GAS AND ELECTRIC CORP.

DOCKET NO. 50-244

FINAL REPORT

DATED: 3 19 84

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SECTION

TABLE, OF CONTENTS

TITLE PAGE

1.0 INTRODUCTION

2.0 IDENTIFIED OVERHEAD HANDLING SYSTEMS

3.0 RESPONSES TO REQUEST FOR INFORMATION INSECTIONS 2.2, 2.3, AND 2.4 of ENCLOSURE 3OF NRC DECEMBER 22 g 1 980 LETTER

4.0 CONCI US IONS

5.0 REFERENCES

ATTACHMENT 1

ATTACHMENT 2

ATTACHMENT 3

ATTACHMENT 4

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..o INTRODUCTION

In response to the NRC Letter of December 22, 1980,"Control of Heavy Loads" this report provides theresults of Rochester Gas and Electric's review of thehandling of heavy loads at our R. E. Ginna NuclearPower Plant.

2.0

Two reports, dated February 1, 1982 and March 2, 1983,have previously been submitted in response to Section2.1 of Enclosure 3 of the December 22, 1980 letter. Inaddition, the NRC issued a Safety Evaluation on thissubject dated January 18, 1984.

This report contains the additional requested informationresponding to Sections 2.2, 2.3 and 2.4 of Enclosure 3of the December 22, 1980 letter. It identifies theextent to which the controls for heavy loads at GinnaStation comply with NUREG-0612 and the recommendationsproposed in order to satisfy the guidelines of NUREG-0612.

IDENTIFIED OVERHEAD HANDLING SYSTEMS

The heavy load overhead handling systems within thescope'f NUREG-0612 were identified in our March 2, 1983submittal and are listed below:

Outside Containment

1. Upper Intermediate Building Monorail

2. 7 1/2 Ton Screenhouse Crane (East)

3. 2 Ton Spent Fuel Handling Crane inAuxiliary Building

4. Monorail in Basement of Auxiliary Building5. Auxiliary Building Overhead Crane with

Main and Auxiliary Hooks

Inside Containment

1. 1 1/2 Ton Fuel Manipulator Bridge

2. 2 Ton Jib at Personnel Hatch

3.

5.

3 Ton Jib at Equipment. Hatch

10 Ton Jib Over Pressurizer Access Hatch

Containment Building Overhead Crane With Main andAuxiliary Hooks

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3.0 RESPONSES TO RE UESTS FOR INFORMATION IN SECTIONS 2.2,2.3 AND 2.4 OF ENCLOSURE 3 OF NRC DECEMBER 22, 1980 LETTER

RESPONSE: The following pages list the guestions andprovide the reguired information to Sections 2.2, 2.3,and 2.4.

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E.2 SPECIFIC RE UIREMENTS FOR OVERHEAD HANDLING SYSTEMSOPERATING IN THE VICINITY OF FUEL STORAGE POOLS

NUREG 0612, Section 5.1.2, provides guidelines concerningthe design and operation of load-handling systems inthe vicinity of stored, spent fuel. Information providedin response to this section should demonstrate thatadequate measures have been taken to ensure that inthis area, either the likelihood of a load drop whichmight damage spent fuel is extremely small, or that theestimated consequences of such a drop will not exceedthe limits set by the evaluation criteria of NUREG0612, Section 5.1, Criteria I through III.

2 ~ 2 1 Identify by name, type, capacity, and equipment designator,any cranes physically capable (i.e., ignoring interlocks,moveable mechanical stops, or operating procedures) ofcarrying loads which could, if dropped, land or fallinto the spent fuel pool.

RESPONSE: There are two overhead handling systems inthe Auxiliary Building which are capable of carryingheavy loads over the spent fuel pool. These cranes area 40 ton Shaw Box overhead crane with a 5 ton auxiliaryhoist and a 2 ton spent fuel bridge crane.

Justify the exclusion of any cranes in this area fromthe above category by verifying that they are incapableof carrying heavy loads or are permanently preventedfrom movement of the hook centerline closer than 15feet. to the pool boundary, or by providing a suitableanalysis demonstrating that for any failure mode, noheavy load can fall into the fuel-storage pool.

RESPONSE: There are no other overhead handling systemszn the vicinity of spent fuel pool. The 40 ton cranemain and auxiliary hooks can move over the spent fuelpool after a series of electrical interlocks have beendefeated. The 2 ton spent fuel pool crane is locatedover the pool. This crane is restricted to loads equalto or less than 1500 lbs. by Ginna Station administrativeprocedures. For further discussion see response toSection 2.2-4.

2 ~ 2 3 Identify any cranes listed in 2.2-1, above, which youhave evaluated as having sufficient design features tomake the likelihood of a load drop extremely small forall loads to be carried and the basis for this evaluation(i.e., complete compliance with NUREG 0612, Section5.1.6 or partial compliance supplemented by suitablealternative or additional design features). For eachcrane so evaluated, provide the load-handling-system(i.e., crane-load-combination) information required.

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RESPONSE: For the two overhead handling systems in thevrcrnity of the fuel storage racks neither has beeneliminated from further evaluation based on Section5.1.6 of NUREG-0612.

2.2-4 For cranes identified in 2.2-1, above, not categorizedaccording to 2.2-3, demonstrate that the criteria ofNUREG 0612, Section 5.1, are satisfied. Compliancewith Criterion IV will be demonstrated in response toSection 2.4 of this request. With respect to Criteria'I through III, provide a discussion of your evaluationof crane operation in the spent fuel area and yourdetermination of compliance. This response shouldinclude the following information for each crane:

a 0 Which alternatives (e.g., 2, 3, or 4) from thoseidentified in NUREG 0612, Section 5.1.2, have beenselected.

b.

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d.

e.

If Alternative 2 or 3 is selected, discuss thecrane motion limitation imposed by electricalinterlocks or mechanical stops and indicate thecircumstances, if any, under which these protectivedevices may be bypassed or removed. Discuss anyadministrative procedures invoked to ensure properauthorization of bypass or removal, and provideany related or proposed technical specification(operational and surveillance) provided to ensurethe operability of such electrical interlocks ormechanical stops.

Where reliance is placed on crane operationallimitations with respect to the time of the storageof certain quantities of spent fuel at specificpost-irradiation decay times, provide presentand/or proposed technical specifications anddiscuss administrative or physical controls providedto ensure that these assumptions remain valid.Where reliance is placed on the physical locationof specific fuel modules at certain post-irradiationdecay times, provide present and/or proposedtechnical specifications and discuss administrativeor physical controls provided to ensure that theseassumptions remain valid.Analyses performed to demonstrate compliance withCriteria I through III should conform to theguidelines of NUREG 0612, Appendix A. Justify anyexception taken to these guidelines, and providethe specific information requested in Attachment2, 3, or 4, as appropriate, for each analysisper formed.

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RESPONSE: (*2 Ton Spent Fuel Bridge Crane) The 2 tonspent fuel pool crane is restricted to lifting fuelassemblies, fuel components such as flow mixers, controlrods and sources, and their associated handling tools.Since this crane only handles spent fuel, and relateditems the crane has been eliminated from further evaluationbased on previous analysis [which satisfies NUREG 0612,Section 5.1.2(4)]. Previous analysis, done in responseto SEP Topic XV-20, (Ref. 8) "Radiological Consequencesof Fuel Damaging Accidents", has shown that the releasesof radioactive material that could result from a fuelhandling accident in this area are well within 10 CFRpart 100 limits. This crane is also prohibited fromlifting loads greater than 1500 lbs. Thus, this cranesatisfies the criteria of NUREG-0612 and need not beconsidered further.RESPONSE: (40 Ton Auxiliary Bldg. Crane Main Hook) Asa result of the potential need to handle spent fuelcasks RGK has decided to modify the main hook of the40 ton auxiliary building crane. This modificationwill allow the main hook to comply with section 5.1.2(1)of NUREG 0612 (single-failure-proof). A description ofthe modification has been submitted to the NRC with alicense amendment request dated January 18, 1984.

The auxiliary 5 ton hook on this crane along with themain hook is restricted from moving over the spent fuelpool by a set of electrical interlocks. The normalhook travel is allowed right to the edge of the pool onthe east end and restricted to 3'-6" from the southedge of the pool. The safe load paths of this craneare shown on Attachment 1.

In addition to the existing electrical interlocks Ginnaadministrative procedures and technical specificationsrestrict the auxiliary building overhead crane frommovement over the spent, fuel pool. The safe load pathsshown on Attachment 1 allow for the movement ofmiscalleneous equipment and material into and out ofthe decontamination area. These loads are generallyless than 1500 lbs. In addition, there are no specificheavy loads which require movement in the area immediatelyadjacent to the fuel pool. The area just east of thepool is the transfer canal and the crane is used herefor inspection of the canal and removing miscellaneousequipment.

* The 2 ton spent, fuel bridge crane consists of a 2 toncapacity bridge with (2) 1 ton hoists attached.

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2.3 SPECIFIC RE UIREMENTS OF OVERHEAD HANDLING SYSTEMS OPERATINGIN THE CONTAINMENT

NUREG 0612, Section 5.1.3, provides guidelines concerningthe design and operation of load-handling systems inthe vicinity of the reactor core. Information providedin response to this section should be sufficient todemonstrate that adequate measures have been taken toensure that in this area, either the likelihood of aload drop which might damage spent fuel is extremelysmall, or that the estimated consequences of such adrop will not exceed the limits set by the evaluationcriteria of NUREG 0612, Section 5.1, Criteria I throughIII.

2.3-1

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Identify by name, type, capacity, and equipment designatorany cranes physically capable (i.e., taking no creditfor any interlocks or operating procedures) of carryingheavy loads over the reactor vessel.

RESPONSE:= There are two overhead handling systemsinside the containment building which are physicallycapable of carrying loads over the reactor vessel.These cranes are a 1 1/2 ton fuel manipulator bridgecrane manufactured by Stearns-Roger Corporation and a100 ton overhead rectangular service crane manufacturedby Whiting Corporation. The rectangular service cranealso has a 20 ton auxiliary hoist.Justify the exclusion of any cranes in this area fromthe above category by verifying that they are incapableof carrying heavy loads, or are permanently preventedfrom the movement of any load either directly over thereactor vessel or to such a location where,. in theevent of any load-handling-system failure, the load mayland in or on the reactor vessel.

2 a 3~3

RESPONSE: The remaining overhead handling systemslocated in the containment building which are withinthe scope of NUREG-0612 are incapable of carrying aheavy load over or near the reactor vessel. Theseoverhead handling systems are listed below:

o 10 Ton Jib Over Pressurizer Access Hatcho 2 Ton Jib at Personnel Hatcho 3 Ton Jib at Equipment Hatch

-Identify any cranes listed in 2.3-1, above, which youhave evaluated as having sufficient design features tomake the likelihood of a load drop extremely small forall loads to be carried and the basis for this evaluation(i.e., complete compliance with NUREG 0612, Section5.1.6, or partial compliance supplemented by suitable

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alternative or additional design features). For eachcrane so evaluated, provide the load-handling-system(i.e., crane-load-combination) information required.

RESPONSE: One crane/heavy load combination has beenelrmrnated from further consideration. This is the 100ton crane load block in the unloaded condition. Thiscrane load block is used for handling numerous heavyloads including those approaching the crane load ratingof 100 tons. In the unloaded condition, as opposed tothe loaded condition, the stresses in the mechanicaland structural components are negligible. Based onthis plus the presence of 2 electrical brakes and onemechanical load brake, the random mechanical failure ofa major load bearing component when moving the crane inthe unloaded condition is not considered credible.

2.3-4 For cranes identified in 2.3-1, above, not categorized.according to 2.3-3, demonstrate that the evaluationcriteria of NUREG 0612, Section 5.1, are satisfied.Compliance with Criterion IV will be demonstrated inyour response to Section 2.4 of this request. Withrespect to Criteria I through III, provide a discussionof your evaluation of crane operation in the containmentand your determination of compliance. This responseshould include the following information for eachcrane:

a ~

b.

Where reliance is placed on the installation anduse of electrical interlocks or mechanical stopsindicate the circumstances under which theseprotective devices can be removed or bypassed andthe administrative procedures .invoked to ensureproper authorization of such action. Discuss anyrelated or proposed technical specification concerningthe bypassing of such interlocks.Where reliance is placed on other, site-specificconsiderations (e.g., refueling sequencing),provide present or proposed technical specificationsand discuss administrative or physical controlsprovided to ensure the continued validity of suchconsiderations.

C. Analyses performed to demonstrate compliance withCriteria I through III should conform with theguidelines of NUREG 0612, Appendix A. Justify anyexception taken to these guidelines, and providethe specific information requested in Attachment2, 3, or 4, as appropriate, for each analysisperformed.

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RESPONSE: RG6cE's Ginna Plant has only two specificloads which must, be carried directly over the reactorvessel, the reactor vessel head and the upper internalspackage. In response to this section load drop analysesin accordance with Appendix A of the NUREG were performed.Based on the modified safe load paths and procedures inplace, the consequences of a drop of the vessel head onthe flange from 16'nd the drop of the upper internalspackage from 15'n the core have been determined to beacceptable. (These values represent the maximum liftheights above the vessel allowed by plant procedures.)The 'Attachments No. 2 and 3 describe the analyses indetail.Ginna Station Administrative Procedure A-1305.1 diagramsthe area of the reactor vessel as a "Restricted Area"when transferring heavy loads. This area is restrictedfor the movement; of all loads with the exception of thereactor head, upper internals package and the reactorcoolant pump and motor parts. In some isolated casesRCF parts are allowed to pass over the very edge of thevessel flange. This load path is necessary because ofthe physical restrictions in the containment buildingand it is limited to the smallest area possible byplant administrative procedures.

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SPECIFIC RE UIREMENTS FOR OVERHEAD HANDLING SYSTEMSOPERATING IN PLANT AREAS CONTAINING E UIPMENT RE UIREDFOR REACTOR SHUTDOWN CORE DECAY HEAT REMOVAL OR SPENTFUEL POOL COOLING

2.4-1

NUREG 0612, Section 5.1.5, provides guidelines concerningthe design and operation of load-handling systems inthe vicinity of equipment or components required forsafe reactor shutdown and decay heat removal. Informationprovided in response to this section should be sufficientto demonstrate that adequate measures have been takento ensure that in these areas, either the likelihood ofa load drop which might prevent safe reactor shutdownor prohibit continued decay heat .removal is extremelysmall, or that damage to such equipment from load dropswill be, limited in order not to result in the loss ofthese safety-related functions. Cranes which must, beevaluated in this section have been previously identifiedin your response to 2.1-1, and their loads in yourresponse to 2.1-3-c.

Identify any cranes listed in 2.1-1, above, which youhave evaluated as having sufficient design features tomike the likelihood of a load drop extremely small forall loads to be carried and the basis for this evaluation(i.e., complete compliance with NUREG 0612, Section5.1.6, or partial compliance supplemented by suitablealternative or additional design features). For eachcrane so evaluated, provide the load-handling-syste'(i.e., crane-load-combination) information specified inAttachment l.

2.4-2

RESPONSE: RG&E has investigated the operation of theupper intermediate building monorail during plantoperation and has restricted the use of this monorailto periods of time when the plant is in a cold shutdownmode. The trolley on the monorail will be locked andwarnings posted. Systems evaluations below the monorailhave shown that a load drop during cold shutdown wouldnot effect any of the required decay heat removalequipment. It is therefore concluded that the use ofthis overhead handling system cannot effect safe shutdownequipment or equipment used for decay heat removal orspent fuel pool cooling.For any cranes identified in 2.1-1 not designated as"single-failure-proof in 2.4-1, a comprehensive hazardevaluation should be provided which includes the followinginformation:

a 0 The presentation in a matrix format of all heavyloads and potential impact areas where damagemight. occur to safety-related equipment. Heavyloads identification should include designation

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and weight or cross-reference to informationprovided in 2.1-3-c. Impact areas should beidentified by construction zones and elevations orby some other method such that the impact area canbe located on the plant general arrangement drawings.

b. For each interaction identified, indicate which ofthe load and impact area combinations can beeliminated because of separation and redundancy ofsafety-related equipment, mechanical stops and/orelectrical interlocks, or other site-specificconsiderations. Elimination on the basis of theaforementioned considerations should be supplementedby the following specific information:

(1) For load/target combinations eliminatedbecause of separation and redundancy ofsafety-related equipment, discuss the basisfor determining that load .drops will not-affect continued system operation (i.e., theability of the system to perform its safety-related function).

(2) Where mechanical stops or electrical interlocksare to be provided, present details showingthe areas where crane travel will be prohibited.Additionally, provide a discussion concerningthe procedures that are to be used for authorizingthe bypassing of interlocks or removablestops, for verifying that interlocks arefunctional prior to crane use, and for verifyingthat'. interlocks are restored to operabilityafter operations which require bypassing havebeen completed.

(3) Where load/target. combinations are eliminatedon the basis of other, site-specific considerations(e.g., maintenance sequencing), providepresent and/or proposed technical specificationsand discuss administrative procedures orphysical constraints invoked to ensure thecontinued validity of such considerations.

RESPONSE: The load/impact matrix sheets for the craneswhrch are located above safety related equipment arecontained in Attachment 4. A discussion of the analysesand results follows:

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7> Ton Screenhouse Crane

The 7 1/2 ton overhead handling crane in the screenhouseis limited to travel west of column line 5 by a lockedtrack switch which prohibits moving the hoist, east ofcolumn line 5. No safety related equipment is locatedwest of column line 5. Use of this crane east ofcolumn line 5 is restricted to the lifts provided inAttachment 4 and transfer of the hoist, into this areais controlled by the Shift Supervisor. The only safetyrelated equipment in this area are the service water(SW) pumps and their associated electrical buses.During normal maintenance operations in this area, onlyone pump would be damaged if a motor were being overhauled.Very infrequently, however, an entire pump assembly maybe pulled. Damage to all four SW pumps, considered avery low likelihood, could be possible under thesecircumstances. Even if it did occur, safe shutdownwould not. be affected. Loss of SW under normal operatingconditions would require a plant shutdown. Even if allservice water were lost, previous evaluations haveshown that safe shutdown can still be achieved anddecay heat can still be removed. [References: NUREG-0821,August 1983, and Appendix R Submittal, dated 1/16/84].

Margin is available in terms of alternative sources ofwater. RGB has installed the capability to providecooling water to the Diesel Generator and AuxiliaryFeedwater System from the yard fire hydrant systems.Procedures are also already in place to cope with aloss of all SW. Thus, although damage to the SW systemis considered a very low probability event, RG&E cancope with this loss through systems and proceduresalready in place at Ginna.

10 Ton Jib Over Pressurizer

This jib is normally used when the plant is at cold orrefueling, shutdown. However, periodic in-serviceinspection requirements have required the use of thisjib during operation to lift and remove one pressurizercavity concrete hatch cover. In order to prohibit apotential load drop into the cavity during operation, amodification to the hatch cover will be installed byJuly 85. The modification will structurally prohibitthe block from falling into the pressurizer cavity. Noother heavy loads are lifted by this jib during operation.

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1'on Fuel Mani ulator Brid e

This crane handles only fuel assemblies and theirassociated handling tools. The analyses done under SEPTopic XV-20 (Reference 8), as in the 2 ton spent fuelpool crane, has shown that the postulated drop of afuel assembly from this crane satisfies section 5.1.2(4)of NUREG 0612.

2 Ton Jib Crane at Personnel Hatch

Systems analyses have been performed on the piping 6equipment below this jib for postulated load drops.The following,equipment could be impacted by a droppedload if the load fell all the way to the basement.

B Accumulator Line to Loop AFan Cooler D and its Service Water Lines

The consequences of a load drop during various modes ofshutdown and during plant operation have been determinedacceptable with the criteria listed in NUREG-0612Section 2.4.

Monorail in Basement of Auxiliar BuildinThe effects of load drops from this monorail are stillbeing evaluated.

3 Ton Jib at E i ment Hatch

The effects of load drops from this jib crane are stillbeing evaluated.

40 5 Ton Auxiliar Buildin Overhead Crane

The effects of load drops from this crane are stillbeing evaluated.

100 5 Ton Containment Overhead Crane

The load/impact matrix sheets in Attachment 4 detailthe analyses performed to date. Some additionalevaluations are still being evaluated.

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c ~ For interactions not eliminated by the analysis of2.4-2-b, above, identify any handling systems forspecific loads which you have evaluated as havingsufficient design features to make the likelihoodof a load drop extremely small and the basis forthis evaluation (i.e., complete compliance withNUREG 0612, Section 5.1.6, or partial compliancesupplemented„by suitable alternative or additionaldesign features). For each crane so evaluated,provide the load-handling-system (i.e., crane-load-combination) information required.

RESPONSE: As described in response to NRC Section2.3-3, the crane load block for the 100 ton containmentoverhead crane in the unloaded condition, has beeneliminated based on sufficient. design factors of safety.d. For interactions not eliminated 'in 2.4-2(b) or

2.4-2(c), above, demonstrate using appropriateanalysis that damage would not preclude operationof sufficient equipment to allow the system toperform its safety function following a load drop(NUREG 0612, Section 5.1, Criterion IV). For eachanalysis so conducted, the following informationshould be provided:

(1) An indication of whether or not, for thespecific load being investigated, the overhead

,crane-handling system is designed and constructedsuch that the hoisting system will retain itsload in the-event of seismic accelerations ',

equivalent to those of a safe shutdown earthquake(SSE).

(2) The basis for any exceptions taken to theanalytical guidelines of NUREG 0612, Appendix A.

(3) The information requested in Attachment 4 ofReference No. l.

RESPONSE: Systems evaluations and justification forthe operation of equipment has been addressed in responseto Section 2.4-2(b).

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CONCLUSIONS

Structural analyses and systems evaluations have shownthat the following cranes at. Ginna Station operatesafely and that the consequences of a load drop havebeen determined to be acceptable using the criteria setforth in NUREG 0612:

1. Upper Intermediate Building Monorail

2. 7Q Ton Screenhouse Crane (East)

3; 2 Ton Spent Fuel Handling Crane in AuxiliaryBuilding

4. 1', Ton Fuel Manipulator Bridge

5. 2 Ton Jib at Equipment Hatch

6. 10 Ton Jib Over Pressurizer Access Hatch (withmodification installed)

7. .Containment Overhead Crane Lifting the ReactorVessel Head and Upper Internals

8. Containment Overhead Crane Load Block in theUnloaded Condition

Additional evaluations are being performed on thefollowing cranes and their potential load drops:

1. Monorail in Basement of Auxiliary Building2. Auxiliary Building Overhead Crane with Main and

Auxiliary Hooks

3. 3 Ton Jib at Equipment Hatch

4. Additional Load Paths for Containment BuildingOverhead Crane

The results of these evaluations will be sent in as anaddition to this submittal by June 29, 1984.

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» 5.0 REFERENCES

NRC Generic Letter "To all Licensees of OperatingPlants and Applicants for Operating Licenses andHolders of Construction Permits" on the Control ofHeavy Loads, December 22, 1980.

2. NUREG-0612, Control of Heavy Loads at NuclearPower Plants, USNRC, July 1980.

3. RG&E Submittal in Response to Reference No. 1,dated February 1, 1982.

4. RG&E Submittal in Response to Reference No. 1,dated March 2, 1983.

.,5. RG&E Submittal in Response to Reference No. 1,dated October 12, 1983.

6. NRC Draft Technical Evaluation Report, datedAugust 19, 1982.

7. NRC Safety Evaluation Report, Phase I, datedJanuary 18, 1984.

8. Letter to John Maier, RG&E, from Dennis M. Crutchfield,NRC, on SEP Topic XU-20, "Fuel Handling AccidentInside Containment", dated October 7, 1981.

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ATTACHMENT 1

Auxiliary Building Crane Travel

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ATTACHMENT 2

Reactor Head Drop Analysis

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Technical Report No. 83084-1Revision 1

FINAL REPORT

FOR THE

REACTOR HEAD DROP ANALYSIS FOR GINNA STATION

JOB NO. 83084

RE E. GINNA NUCLEAR POWER STATION

ROCHESTER GAS AND ELECTRIC CORPORATIONROCHESTER, NEM YORK

COPY NO. C33

PreparedLe d Engineer

ApproveP oject Manager

Approved NQJ&i A 3 wn .5//e/8'fRochester Gas 8 Electric

Cygna Ener gy Ser vices600 Atlantic Avenue

Boston, Massachusetts 02210

TABLE OF CONTENTS

Section

1.0 Introduction

2.0 Reactor Vessel Description

3.0 Problem Description

4.0 Acceptance Criteria5.0 Analysis Methodology and Results

5.1 Basic Finite Element Model5.2 Uniform Drop Analysis5.3 Concentrated Drop Analysis

6.0 Conclusions and Recommendations

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A endices

ReferencesFigures 1 through 51

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FINAL REPORT

Reactor Head Dro Anal sis

1.0 INTRODUCTION

This report documents the results of Cygna 's pilot study analysisof the reactor vessel head dropping onto the reactor vessel atGinna. The study is intended to form the basis for a response toissues raised in NUREG-0612.

Two basic analyses were conducted. The first was an elastictime-history analysis of the reactor head dropping uniformly ontothe vessel flange. The second was a plastic time-history analysisof the reactor head dropping on edge (concentrated drop) onto thevessel flange directly over a primary coolant nozzle. The studyconcludes that core cooling can be maintained in the unlikelyevent of either a uniform drop from fifteen feet or a con-centrated drop from sixteen feet.

2 ~ 0 REACTOR VESSEL DESCRIPTION

The Reactor Pressure Vessel (RPV) at Ginna consists of five basiccomponents. These components include (1) a hemispherical closureassembly (RPV head), (2) a flange forging to receive the closurehead assembly, (3) an upper shell region containing the low headsafety injection and primary coolant inlet and outlet nozzles,(4) an intermediate shell region, and (5) a bottom hemisphericalshell containing instrumentation nozzles (Figure 1). The vessel

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is approximately 13 feet in diameter and is supported at sixpoints by the four reactor coolant loop nozzles plus two lugs

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~which all rest on support pads. The six supports are locatedapproximately 8'-6" below the vessel flange and are spacedcircumfere'ntially at 60'ntervals (Figures 2 & 3).

The reactor vessel flange is a machined for ging welded to theupper shell. The 14.5" thick flange is fabricated from SA-336manganese-molybdenum steel conforming to ASME Code Case 1332-1and is clad internally with weld deposited austenitic stainlesssteel.

The upper shell of the vessel is a machined forging welded to thereactor vessel flange and to the intermediate shell. The 9.0"thick upper shell is also fabricated from SA-336 manganese-molybdenum steel conforming to ASME Code Case 1332-1 and is cladinternally with weld deposited stainless steel. The upper shellcontains the four primary coolant nozzles, the vessel supports,and the two safety injection nozzles. The two 27.5" I.D. inletnozzles and the two 29.0" I.D. outlet nozzles are approximately10" thick. These nozzle forgings are fabricated from the same

material as the upper shell and are also clad internally withweld deposited austenitic stainless steel. The two safetyinjection nozzles are welded into the upper shell. These nozzlesinject water into the vessel in the event of a primary coolantfailure. The centerline of both the safety injection nozzles andthe primary coolant nozzles are located approximately 6'-2" belowthe top surface of the vessel flange.

The upper shell is welded to the intermediate shell which isfabricated from A-508 Class 2 manganese-molybdenum steel. The

intermediate shell is 6.5" thick and is internally clad with welddeposited austenitic stainless steel.

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The bottom hemispherical head is welded to the intermediate shellof the vessel. The 4.1" thick hemispherical head is forged froma single plate of SA-302 Grade B manganese-molybdenum steel and

is clad internally with weld deposited austenitic stainless steel.

3.0 PROBLEM DESCRIPTIOH

During refueling operations the RPV head, which weighs 134,550pounds, is lifted vertically from the vessel flange by the 100

ton containment overhead crane. Vertical alignment is maintainedduring part of this lift by three guide studs approximately 15

feet high which are engaged in the vessel flange (Figures 4 & 5 ).After clearing the guide studs, the head is lifted for some

additional height (Figure 6 ) before being moved later ally to thesouth (over the refueling cavity floor), and eventually to thehead storage stand on the operating floor.

While the RPV head is over the vessel, two types of drops can

occur. One is a uniform drop onto the RPV flange which is most

likely to occur while the RPV head is on the guide studs. The

other is a concentrated drop onto the RPV flange which can onlyoccur when the RPV head has been lifted above the top of theguide studs.

In the analyses performed by Cygna, the uniform drop was based

upon the head falling 15 feet through air. This is the maximum

drop height for which the head would still be aligned on theguide studs, thus insuring a uniform drop (Figure 7 ). Afterclearing the guide studs, a uniform drop can no longer be

guaranteed. However, at this point, the head. is now free to move

laterally to the south from its position directly over the core.

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For this reason, the concentrated drop was based upon the headfalling 16 feet, since this was the minimum height for which a

concentrated drop scenario could be postulated (Figures 8 & 9).

For both the uniform drop and the concentrated drop it was con-servatively assumed that, in addition to the weight of the RPV

head, the total load dropped onto the RPV flange included theweight of the lifting rig (30,000 pounds) and the crane hookblock (8,500 pounds). Thus, the total load dropped was 173,050pounds.

As required by NUREG-0612, the head was conservatively assumed tobe rigid in all analyses; thus it absor bed no energy. Furthermore,it was assumed that during the concentrated drop the entire loadstruck the RPV flange directly above a primary coolant nozzle,thus imposing the most severe conditions on a single nozzle.

4. 0 ACCEPTANCE CRITEREA

During reactor shutdown, it is essential to provide coolingwater to the reactor core. As a minimum this requires thatthe safety injection nozzles maintain structural integrity aftera RPV head drop event. Thus, the consequences of a head dropover the RPV are acceptable if the analysis conservativelydemonstrates that the structural integrity of the safetyinjection nozzles is maintained after a RPV head drop. There-fore, it is essential that during a head drop event the reactorsupport system be maintained, since this virtually insures theintegrity of the safety injection nozzles.

Unlike the safety injection nozzles, the primary coolant loopnozzles are part of the reactor support system and are directly

~ b

l1

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loaded during the head drop impact . The basis for determiningprimary coolant nozzle integrity was developed from ultimatestrain behavior. The ultimate strain is the point at which a

material will rupture in simple tension. A reasonable acceptancecriteria f'r maintaining primary coolant nozzle integrity is onein which the peak equivalent strain (which is proportional to theoctahedral shear strain) in the vicinity of the vessel/nozzleintersection does not exceed one half of the minimum ultimatestrain of the material. (It should be noted that in uniaxialtension, the uniaxial ultimate strain is equal to the equivalentultimate .strain.) Since the peak eguivalent strain at thenozzle/vessel intersection is a highly localized strain resultingfrom a combination of geometric discontinuity and strain con-centration, limiting the peak equivalent strain to one half ofthe minimum ultimate strain of the material should insure thatthe actual impact kinetic energy is well below the kinetic energylevel that would produce complete plastic instability (collapse )

of the nozzle support.

The material from which the reactor vessel upper shell andcoolant nozzles at Ginna is fabricated has a minimum ultimatestrain capacity equal to 0.17 in/in for material 9.0 inches thick(ASME Code Case 1236-2). Thus, to maintain nozzle integrity, thepeak equivalent strain in the vicinity of the nozzle/vesselintersection shall not exceed 0.085 in/in during a RPV head dropevent. Should this strain level be exceeded, the nozzle wouldnot be considered functional from an analytical point of view;however, reactor core cooling could still be maintained by thesafety injection nozzles, provided the integrity of these nozzlescan be demonstrated.

Under dynamic impact, strain criteria alone is not sufficient todemonstrate material integrity since strain rate effects may also

g P

be important. Therefore, as an additional criteria to conserva-tively demonstrate material integrity, strain rates should be

maintained at a level below 100 in/in/sec.

It is assumed that material flaws are small, and that at the timeof impact the vessel material is at a temperature which is equalto or greater than the nil ductility temperature plus sixtydegrees fahrenheit.

5.0 ANALYSIS METHODOLOGY AND RESULTS

5.1 Basic Finite Element Model

The basic finite element model for both the uniform and concentrateddrop analysis consists of a 30'ircumferential segment of theRPV cutting directly through the center line of a primary coolantloop nozzle (Figure 10 & 11). The basic model was developed on

the ANSYS computer program (Version 4.1B) and consists of (1) thevessel flange, (2) the upper shell region, (3) a primary coolantloop nozzle, (4) a vessel support, and (5) an extended inter-mediate shell which incorporates an additional length to compen-sate for the hemispher ical bottom. By properly specifying theshell boundary conditions at the edges of the 30'egment, thisfinite element model can represent the entire 360 degrees of thevessel, provided that (1) one assumes that the actual vessel hassix nozzle supports instead of four (i.e., two nozzle supportsreplace the two lug supports ), and (2) the loads are uniformlyapplied to the entire vessel. Thus, the 30'egment model givesalmost an exact solution for the case of a uniform head drop.For the concentrated drop analysis, where a single impact occursdirectly above one nozzle, the 30'egment model is very conser-

vative since the symmetric boundary conditions imply that thesame load is applied to all twelve of the 30'egments comprisingthe full vessel. Thus, in the concentrated drop analysis using a

30'egment finite element model, a single nozzle (60'egment ofthe vessel') will resist the impact of the entire head drop.

The 30'egment model is the minimum size finite element model

which could be used for the studies. It was intended that thismodel would produce results which would form the basis for makinga decision as to whether to pur sue a more refined analyticalresolution to the problem using a more elaborate finite elementmodel, or to attempt a different approach, such as a probabilistic(PRA) resolution to the head drop problem.

The basic geometry of the finite element mesh for the 30'egmentmodel was determined from the fundamental cylindrical shell para-meter ~Rt, where R is the mean shell radius and t is the shellthickness. This parameter determines the magnitude of thebending boundary length and is a measure of the rate at which a

uniformly applied circumferential'isturbance "damps out" alongthe meridional dir ection of'he shell. For the reactor vessel,this shell parameter is equal to 25 inches, and for the primarycoolant nozzle it is equal to 14 inches. From Cygna'sexperience, when using simple shell elements, elements no largerthan approximately 0.4 ~Rt should be used within the regionswhere the highest bendings stress gradients are expected.Consequently, for the reactor vessel shell, the minimum elementsize selected was 8 inches, while for the nozzle a 6 inch minimumsize was used.

In the circumferential direction of the vessel, the elementlength encompasses approximately 10'f arc, while in the nozzle

22.5'f arc was used. The 22.5'f arc was a compromise between

efficiency and accuracy but was considered adequate for theaccuracy required by the studies, since a model with 15'f arc(the "accepted" maximum) would have increased the model size by

50'S. (As 'will be discussed in Section 5.3 of this report, thisare size was eventually reduced to 11.75'n the most highlystressed regions of the nozzle.)

Thin shell finite elements were used for both elastic and plasticanalysis. Overall, more detailed behavior could have been

achieved using thick shell elements, however, this would have

required using three elements through the shell thickness whichwould have tripled the size of the basic model and would havemade model development, analysis, and post-processing iar tootime consuming and expensive for pilot studies. Thus, from a

cost, efficiency, and schedule point of view, the purpose of thepilot studies dictated that thin shell elements be used. From

the standpoint of accuracy, the thin shell elements will error on

the conservative side since through-thickness shear deformationsare neglected in thin shell elements. This means that impactenergy which would have been absorbed in shear deformation mustnow be expended in bending deformation, resulting in far highersurface normal stresses and strains than would have existed inthe thick shell elements.

Initially, three thin shell quadrilateral finite element mesh

configurations were developed and tested statically to determinethe influence of certain mesh configurations on cost and accuracy.All three meshes gave comparable results which indicated that themodel was stable and essentially independent of these configura-tion changes.

7

~~ >a

The nozzle supports were modeled with two quadrilateral elementswhich had an equivalent thickness calculated to duplicate theequivalent axial stiffness of the support from the nozzle to theconcrete support structure.

5.2 Uniform Drop Analysis

Descri tion

The uniform drop analysis was performed on the basic finiteelement model described in Section 5.1. This model (30'ircum-ferential segment) included one nozzle and its supporting pad,and was made from quadrilateral shell elements. The element andnode numbers of this model for the upper shell, intermediateshell and primary coolant nozzle are shown in Figures 12 thru 19.To insure that the model was behaving properly, a static analysiswas performed.

To simulate the uniform impact of the RPV head with the vesselflange four gap finite elements were used. These gap elementsconnect nodes 6 through 9 to nodes 10 through 13, respectively(Figure 13). A proportioned mass of the RPV head was placed atnodes 6 through 9, and the finite gap between the RPV head (nodes6, 7, 8 8 9) and the RPV flange (nodes 10, 11, 12 8 13) was madeto close within a prescribed time step so that the head wouldimpact the flange at precisely 373 inches/second (i.e., thecalculated velocity dropping through air from 15 feet).

An elastic dynamic impact analysis was then performed for atotal duration of 5.0 milliseconds using a time step 0.010 milli-seconds. This small time step was required to resolve theresponse of the vessel at the point of impact of the reactor head

V

and to insure an accurate representation of the propagation ofthe compression wave through the smallest element of the vesselshell.

As required by NUREG-0612, the head was conservatively assumed tobe rigid during impact; thus it absorbed no energy. Xn addition,the head mass also included the mass of the lifting device and

hook block. All elements were assumed to remain elastic for theuniform dr op.

Discussion of Results

When the RPV head is dropped from 15 feet above the vessel, theuniform impact of the head with the flange initially producesvery high contact forces of extremely short duration (Figure 20).As the flange begins to displace downward (Figure 21) thecompression stress wave pr opagates through the vessel shell,eventually. reaching and displacing the bottom of the vessel as itis reflected from the bottom edge (Figure 22). During thistime, high reaction forces develop at the nozzle support (Figure23). These high reaction forces r otate and deform the nozzle toproduce high stresses in the bottom of the nozzle at the nozzle/vessel intersection. The maximum equivalent stress anywhere inthe vessel model occurs at this location (element 60, node 35)(Figures 13, 16, 17). The magnitude of this maximum equivalentstress is approximately 200,000 psi, which is well above thematerial yield point of 50,000 psi, and occurs at 2.36 milli-seconds after impact (Figure 24). (The equivalent stress, or

Von Mises str ess, is used to identify the onset of yielding ina material subjected to a general state of stress. When theequivalent stress for a general state of stress exceeds the yieldpoint in simple tension, the material yields in the general state

t

I J

I't>t,-

i'

of stress.) The RPV flange does not yield during impact and, as

might be expected, general yielding is confined to the nozzle and

upper shell within the vicinity of the nozzle/vessel intersection.

It is not possible from this elastic analysis to determine themagnitude of the maximum plastic strains which will occur in thevessel during a uniform drop of the RPV head because the vesselis highly indeterminate. As a consequence, the initiation ofplasticity within the vessel will lead to a complex redistribu-tion of internal forces and an overall softening of the dynamicresponse. However the maximum lastic res onse of the vesseldurin a uniform dr o is bounded b the results of the concen-trated dro anal sis since the total im act kinetic ener ofthe reactor head in the concentrated dro anal sis is six timesthe ener in ut to the uniform dro anal sis.

5.3 Concentrated Drop Analysis

The initial concentrated drop analyses were performed on themodel developed for the uniform drop. Using this model, a dy-namic analysis was performed to determine the extent to which theelastic quadrilateral elements exceeded the minimum elastic limitwhen subjected to a heavy concentrated drop. Those elementsidentified as potentially capable of developing plastic strainswere modified to triangular plasticity elements as required bythe element library of the ANSYS program. On the basis ofadditional analyses, this model was further refined, eventuallyevolving into the final model shown in Figures 25 thru 38.

-11-

t

To simulate the concentrated impact of the RPV head with thevessel flange, a single gap finite element connecting nodes 6 and10, was used (Figure 31). Since the 30'egment finite elementmodel constitutes half of a single nozzle segment of the reactor,only one half of the total mass of the RPV head was placed atnode 6. The finite gap between the RPV head (node 6) and the RPV

flange (node 10) was made to close within a prescribed time stepso that the head would impact the flange at precisely 385inches/second (Figure 5 1), thus simulating a drop in air of thecenter of mass of the RPV head through 16 feet.

Using a time step of 0.010 milliseconds, the nonlinear dynamicanalysis was then performed for a total duration of 13.0 milli-seconds. Lower bound material properties were conservativelyassumed with no increase in strength taken for dynamic effects.To insure that maximum energy was absorbed by the nozzle andvessel, the support pad beneath the nozzle was required to remainelastic during all plasticity analyses. As before, the head was

~

~

~

conservatively assumed to be rigid during impact; thus itabsorbed no energy. In addition, the head mass also included themass of lifting device and hook block.

Based on the results of this nonlinear dynamic analysis, itbecame evident that the likelihood was high that the r eactorcould maintain core cooling after a concentrated head drop from16 feet, and that a PRA approach would be unnecessary. However,the results from the pilot study models were not refined enoughin the most highly stressed regions of the vessel to effectivelydetermine the magnitude of the highest strains. Therefore, Cygnainitiated a number of model changes to further refine the finiteelement mesh in the highly strained tensile region of the vessel/nozzle intersection to insur e that the high strain gradients in

r

P

this region were not missed in the analysis. This refinement~

~

resulted in the ratio of the smallest element dimension in thedirection of the strain gr adient to the ~Rt being equal to 0.2.A nonlinear dynamic analysis was then performed with the refinedmodel using the same time step and duration. The intent of thisanalysis was to demonstrate that the vessel could inherently

t

absorb sufficient strain energy through plastic deformationwithout jeopardizing the structural integrity of a single nozzleunder the most adverse conditions and with the most conservativeassumptions possible.

Discussion of Results

When the RPV head is dropped from 16 feet above the vessel, theconcentrated impact of the head with the flange at a pointdirectly above the nozzle (node 10) (Figure 31) produces a nearlya constant contact force of 10 million pounds for a duration ofapproximately 11 milliseconds (Figure 39). As the flange beginsto displace downward (Figure 40), the compressive stress wavepr opagates through the vessel shell, eventually reaching anddisplacing the bottom of the vessel (Figure 41). In a localregion directly under the point of contact, large amounts ofplastic deformation occur, and a considerable amount of strainenergy is absorbed. In fact, 52 percent of the initial kineticenergy of the RPV head (Figure 42) is absorbed as strain energywithin elements 4 and 5 (Figures 43 E 44). The average strainenergy density within these two elements is 2500 in-lbs/cubicinch. The local nature of this plastic deformation is obviousfrom Figure 40 which shows the displacement of the nodes at thetop of the flange. The local nature of this high plasticity isfurther illustrated by the fact that the total strain energyabsorbed by all of the adjacent elements (elements 6, 7, 8, 9,

-13-

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„~~ 5

10 8 24) (Figure 29) amounts to only 10 percent of the initialkinetic energy.

The maximum reaction within the nozzle support pad is slightlygreater than 16 million pounds and occurs at 5.5 millisecondsafter impact (Figure 45). This reaction is quite high; neverthe-less, the support stresses remain within the elastic range, whichis consistent with the assumption made in the nonlinear dynamicmodel.

The most highly stressed tensile region of the reactor vessel isin the bottom of the nozzle at the intersection of the nozzle and

vessel shell. The maximum equivalent strain at the intersectionis 0.044 in/in and occurs within element 106 at 7.0 millisecondsafter impact (Figure 46). This maximum strain is well below theacceptance strain level for nozzle integrity by a factor of 1.9.However, due to the presence of the geometric discontinuity atthe intersection of the nozzle and vessel shell, a strain con-centration will occur . This in turn will increase the maximum

equivalent strain at the intersection. Based on the ratio ofnozzle fillet radius to shell thickness, the strain concentrationfactor is approximately 1.4. An estimate of the peak equivalentstrain can be made by multiplying the maximum equivalent strainby the strain concentration factor. This results in an estimatedpeak equivalent strain of 0.062 in/in which is well below theacceptance strain for nozzle integrity by a factor of 1.37. The

peak strain rate of 22 in/in/second (15.5 x 1.4) occurs at 2.3milliseconds after impact (Figure 47) and is well below the accep-tance value for nozzle integrity. Therefore, the nozzle remainsintact.

The two safety injection nozzles are each located adjacent to a

primary coolant inlet nozzle. Their location corresponds closely

$t

IV

I

with node 34 (Figure 31) on the finite element model. The

vertical and radial displacements and rotation of this node areshown in Figures 48, 49 and 50. Based upon the magnitude of thesedisplacements and the layout of the safety injection piping, thestresses in the safety injection piping should remain within theelastic range during a concentrated head drop event, therebyinsuring the integrity of the safety injection system.

It has already been noted that the energy absorbed in plasticdeformation at the point of impact of the reactor head with theflange is approximately 52% of the total kinetic energy of thereactor head at the moment of impact. In the unlikely event ofan actual head drop, much of the remaining kinetic energy, whichhas been assumed in the analysis to be totally absorbed by a

single nozzle and support, would be absorbed by the adjacentnozzle on one side and lug support on the other. (A simplestatic calculation shows that for a concentrated force applieddirectly to the flange above a nozzle, 56% of this force isreacted by the nozzle support directly beneath the applied forceand 22% is reacted by each of the adjacent nozzle/lug supports.)In the actual situation, when the nozzle directly under the pointof impact begins to absorb energy in plastic deformation, theadjacent nozzle and lug support would be absorbing energyelastically. The plastic "softening" of the nozzle would resultin a significant redistribution of forces from the plasticallydeforming nozzle to the adjacent nozzle and lug support. As thisredistribution continues, the adjacent nozzle and lug supportwould begin to deform plastically and absorb large amounts ofenergy. In addition, it is important to note that the mass ofthe resisting structure (30'egment model) grossly underestimatesthe effective mass of the actual vessel that would be mobilizedduring the impact. By underestimating the active mass of the

-15-

r

I

vessel, the inertial resistance of the vessel is reduced andvessel response (stresses, strains and displacements ) is increased.Therefore, since all of the energy which could be absorbed by theadjacent supports has been neglected, and since the resistingmass of the adjacent portions of the reactor has been neglected,and in view of the numerous other conservative assumptionsalready mentioned, the calculated peak equivalent strain of 0.062in/in is considered a ~ver conservative upper bound.

6. 0 CONCLUSION AND RECOMMENDATIONS

Cygna's analyses have conservatively demonstrated that the RPV

head can be dropped through air from a clear height of 16 feetabove the vessel flange while still maintaining the integrity ofthe vessel support system. With the vessel support system intact,core cooling is maintained, since, as an absolute minimum, theintegrity of the safety injection nozzles is assured.

Based on the results of these analyses, the reactor pressurevessel can withstand the consequences of a postulated head dropfrom a height of 16 feet. However, the present operating proce-dures call for the head to be lifted to an elevation above theoperating floor level (i.e., a lifted height greater than 23feet ) before moving to the south over the refueling cavity floorar ea.

Cygna recommends that the RPV head lifting procedures be modifiedslightly to allow the head to move later ally to the south as soonas the head safely clears the guide studs. The head should bemoved sufficiently to the south to completely .clear the area overthe vessel before proceeding vertically again. This recommendationis made for the following reasons:

-16-

1

1. The vessel can withstand a postulated head drop under thisrevised procedure.

2. Moving the head away from the area directly over the core ismost c'bnsistent with the NRC's "defense in depth" philosophywith regards to NUREG-0612.

3. The horizontal load path traversed by this revised procedureis the same as for the current procedure. The head would notbe lifted over any different area.

AP P ENDIX A

References

REFERENCES

1. Instruct'ion Manual - Reactor Pressure Vessel, RGE-105, The

Babcock and Wilcox Company, Contract No. 610-0110.

2. ANSYS Engineering Analysis System User's Manual, Revision4.1, Swanson Analysis System, Inc., Houston, PA.

3. ANSYS Engineering Analysis System Theoretical Manual,November 1977, Swanson Analysis Systems, Inc., Houston, PA.

4. Theory and Design of Modern Pressure Vessels, by John F.Harvey, Second Edition, Van Nostrand Reinhold Co., 1974.

5. Roar k, Hartenberg and Williams: "Influence of Form and Scaleon Strength", Engineering Experiment Station, University ofWisconsin Bulletin, 1938.

s. Gilber t Associates, Inc., Drawing Nos. D-421-018, Rev. 5;D-421-051, Rev. 0; D-421-052, Rev. 0; D-521-044, Rev. 10 and

D-521-046, Rev. 5.

7. Babcock and Wilcox Co., Drawing Nos. 117825E, Rev. 5,117820E, Rev. 5.

8. Letter: M. Fitzsimmons (RG&E) to J. McWilliam (Cygna }

8/10/83.

9. Letter: M. Fitzsimmons (RG&E) to J. McWilliam (Cygna)8/12/83.

10. Letter: M. Fitzsimmons (RG&E) to J. McWilliam (Cygna)8/22/82.

APPENDIX B

Figures 1 through 51

r

I

LIST OF FIGURES

1. Reactor. Pressure Vessel Cross Section

2. Reactor Pressur e Vessel - Longtudinal Section Dimensions

3. Reactor Pressure Vessel — Horizontal Section through Nozzles

4. Reactor Vessel with Reactor Head Attached

5. Reactor Head at Top of Guide Studs

6. Reactor Head Decoupled from Guide Studs

7. Reactor Head Uniform Drop

8. Nonuniform Drop of Reactor Head

9. Reactor Head Concentrated Drop - Impact Directly above

Nozzle

10. ANSYS Finite Element Model for Uniform Drop Analysis - SideView of Complete Model

11. ANSYS Finite Element Model for Uniform Drop Analysis — 3D

Hidden Line Plot of Complete Model

12. Element Numbers for the Flange and Upper Shell Portions ofthe Uniform Drop Model

13. Node Numbers for the Flange and Upper Shell Portions of theUniform Drop Model

14. Element. Numbers for the Extended Intermediate Shell Portionof. the Uniform Drop Model

15. Node Numbers for the Extended Intermediate Shell Portion ofthe Uniform Drop Model

16. Nozzle Element Numbers for 0he Uniform Drop Model

17. Nozzle Node Numbers for the Uniform Drop Model

18. Support Saddle Element Numbers for the Uniform Drop Model

19. Support Saddle Node Numbers for the Uniform Drop Model

20. Contact Force (lbs) between the Reactor Head and VesselFlange for a Uniform Drop versus Time (seconds )

21. Displacement (inches) of Nodes 10, 11, 12 and 13 at the Topof the Reactor Vessel Flange for a Uniform Drop versus Time(seconds)

22. Displacement (inches ) of Node 61 at the Bottom of the Vesselfor a Uniform Drop versus Time (seconds )

23. Reaction Force (lbs) at the Nozzle Support for a UniformDrop versus Time (seconds )

24. Equivalent Stress (psi ) in Element 60 at Node 35 versus Time(seconds )

-B2-

25. ANSYS Finite Element Model for Concentrated Drop Analysis3D Hidden Line Plot of Complete Model

26. ANSYS Finite Element Model for Concentrated Drop Analysis-3D Hidden Line Plot of the Vessel Flange and Upper Shell

27. ANSYS Finite Element Model for Concentrated Drop Analysis—Side View of Complete Model

28. ANSYS Finite Element Model for Concentrated Drop Analysis-Front View of Complete Model

29. Element Numbers for the Flange and Upper Shell Portions ofthe Concentrated Drop Model

30. Element Numbers for the Lower Half of the Upper ShellPortion of the Concentrated Drop Model

31. Node Numbers for the Flange and Upper Shell Portions of theConcentrated Drop Model

32. Node Numbers for the Lower Half of the Upper Shell Portionof the Concentrated Drop Model

33. Element Numbers for the Extended Intermediate Shell Portionof the Concentrated Drop Model

34. Node Numbers for the Extended Intermediate Shell Portion ofthe Concentrated Drop Model

35. Nozzle Element Numbers for the Concentrated Drop ModelSide View

-B3-

'C

36. Element Numbers for the Lower Half of the Nozzle - Top View,Concentrated Drop Model

37. Nozzle Node Numbers for the Concentrated Drop Model — SideView

38. Node Numbers for the Lower Half of the Nozzle — Top View,Concentrated Drop Model

39. Contact Force (lbs) between the Reactor Head and VesselFlange for a Concentrated Drop versus Time (seconds)

40. Displacement (inches) of Nodes 10, 11,'2 and 13 at the Topof the Reactor Vessel Flange for a Concentrated Drop versusTime (seconds)

41. Displacement (inches ) of Node 61 at the Bottom of the Vesselfor a Concentrated Drop versus Time (seconds)

42. Kinetic Energy (inch-pounds ) of the Reactor Head versus Time(seconds )

43. Strain Energy (inch-pounds) Absorb bed by Element 4 versusTime (seconds)

44. Strain Energy (inch-pounds ) absorbed by Element 5 versusTime (seconds )

45. Reaction Force (lbs) at the Nozzle Support for aConcentrated Drop versus Time (seconds )

46. Equivalent Strain at the Top (Positive Face) and Bottom(Negative Face) of Element 106 for a Concentrated Dropversus Time (seconds)

-B4-

47. Equivalent Strain Rate (in/in/sec) at the Top (PositiveFace ) of Element 106 for a Concentrated Drop versus Time(seconds)

48. Vertical Displacement (inches ) of Node 34, which is theapproximate location of the Safety Injection Nozzle,versus Time (seconds )

49. Radial Displacement (inches ) of Node 34, which is theapproximate location of the Safety Injection Nozzle, versusTime (seconds).

50. Rotation (radians) about the Circumferential Axis of Node

34, which is the Approximate Location of the SafetyInjection Nozzle, versus Time (seconds )

51. Reactor Head Velocity (inches/second, Node 6) versus Time(seconds )

-B5-

CONTROL ROD MECHANISMHOU5ING (MK-A33 THRU MKA40)

SHROUD SUPPORT A FLANGE(MK-56 THRU MK 60)

LIFTING LUG (MK-30)

g* n

I

VENT PIPE (MK.6I)(REF. INSULATION)

CLOSURE STUD (MK.4I)

NUT (MK-42)RPV HEAD

CLOSURE HEAD

I

MATING

ISURFACE

I %. ~

SP)CRICAL WA5HERS(MK.43 A MK-44)

FLANGE

UPPERSHELL

REFUELING SEALLEDGE (MK-I5)

rMONITORING TUBE(MK-I9 A MK-20)

OUTLET NOZZLE(MK (3)

VESSEL SUPPORTWELD PAD

CORE SUPPOR LEDGE

.FgL

OUTER 0-RINGGASKET (MK-25)

INNER O.RINGGASKET (MK-24)

I)4.ET NOZZLE'(MK-l2)

VESSEI. SUPPORTWELD PAD

INTERMEDIATESHELL

VESSEL SUPPORT(MK-l6 A MK-I 7 )

SAFETY INJECTIONNOZZI.E <MK.B)

CORE SUPPORTGUIDE (MK-7) REACTOR VESSEL

BOTTOMSHELL

NOZZLES(MK.65 THRU MK.89)

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'6

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/~

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1

Sl/PPORT P(d / /~de( P(PP(//(r(l/4IR ~ ~ e pIOl./OR A'O((le /MC

X""0'T

g-''i

P/PP/l

FIGURE 3: Reactor Pressure Vessel — Horizontal Sectionthrough Nozzles

~ ~

~ ~ ~

i?

X4Vh

FIGURE 4: Reactor Vessel with Reactor Head Attached

S ~

GUIDESTUDS S

0I

SIl

~ ~ ~

FlGURE 5: Reactor Head at Top of Guide Studs

GUIDESTUDS~ O OI

FIGURE 6: Reactor Head Decoupled from Guide Studs

t

r

0

FIGURE 7: Reactor Head Uniform Drop

)'A

~ ~ ~

FIGURE 8: Nonuniform Drop of Reactor Head

FIGURE 9: Reactor Head Concentrated Drop — ImpactDirectly above Nozzle

AGEE REACTOR PRESSURE VESSEL HOOEL. 30 OEGREE SEGHENT (SUE)

FIGURE 10: ANSYS Finite Element Model for Uniform DropAnalysis — Side View of Complete Model

r

RGCE RERCTDA PRESSUAE VESSEL -- UNIFORM OROP. IS FT, ELASTIC

Figure ll: ANSYS Finite Element Model for Uniform DropAnalysis — 3D Hidden Line Plot of Complete Model

pr

13 14 15

21 22 23

2028

29

1927

33

Se

e256

RGCE REACTOR PRESSURE VESSEL HOOEL, 30 DEGREE SEGHENT (SUE)

FIGURE 12: Element Numbers for the Flange and Upper ShellPortions of the Uniform Drop Model

6 7 8 9

0 1 2 3

19 15 1e 17

18 19 20 2l

22 23

2% 25

3Q 31

2~

r

32 33 39

SB/ ~ 3e '9O

.w ~/93 9'l

81 'M2

95 Re 97 9eRGCE RERCTGR PRESSURE VESSEL HODEL. 30 DEGREE SEGMENT (SUE)

t

FIGURE 13: Node Numbers for the Flange and Upper Shell Portionsof the Uniform Drop Model

63 65

66 67

ee ee

70 71

72 73

AGEE REACTOR PRESSURE VESSEL HOOEL. 30 OEGREE SEGNENT tSUE1

FIGURE 14: Element Numbers for the Extended Intermediate ShellPortion of the Uniform Drop Model

52 53 59

55 56 57

58 59 60

eL ea e3AG4E REACTOR PRESSURE VESSEL MODEL, 30 DEGREE SEGMENT tSUE)

FIGURE 15: Node Numbers for the Extended Intermediate ShellPortion of the Uniform Drop Model

kf* 4

ie ]7 18

30 31 32

35 36 37

90

50 52

57

RGCE REACTOR PRESSURE VESSEL MODEL, 30 OEGAEE SEGHENT (SUE)

59 eo

FIGURE 16: Nozzle Element Numbers for the Uniform Drop Model

72; 81

10

90 26

70 79 BB 28

69 78 87 29

77 32

76 38

66 75

6S 79

I3RGCE REACTOR PRESSURE VESSEL H00EL. 30 OEGREE SEGHENT (SUE)

83 36

FIGURE 17: Nozzle Node Numbers for the Uniform Drop Model

58

S1

AGCE RERCTDR PRESSU8E VESSEL ROOEL, 30 OEGREE SEGMENT (SUE)

FIGURE 18: Support Saddle Element Numbers for the Uniform Drop l1odel

g

75

3RG4E REACTOR PRESSURE VESSEL HOOEL. 30 OEGREE SEGMENT ISUE)

FIGURE 19: Support Saddle Node Numbers for the Uniform Drop Model

FORC

NPACT

-90

-00

-120

-Ieo

-200

-2'10

-280

-320

-360 INE

0 .ODOSO .00100 .00150 .00200 .0025D .00300 .00350 .00'400 .00950 ~ ODSOO

ROLE REAC'IOR PRESSURE VESSEL -- UNIFORN DROP, 15 FTo ELRSTIC

FIGURE 201 Contact Force (lbs) between the Reactor Head and VesselFlange for a Uniform Drop versus Time (seconds)

o12

DISP

.08

e 0'4

.04

-. 08

-. I2

-. Ie

-o20

.24

-. 28IHE

.00050 oOOIDO .OOISO .00200 .00250 .00300 .00350 .00400 .00450 .00500

ACUTE REACTOR PRESSURE VESSEL -- UNIFORM DROP. IS FT, ELASTICP

FIGURE 21: Displacement (inches) of Nodes 10, ll, 12 and 13 at theTop of the Reactor Vessel Flange for a Uniform Dropversus Time (seconds)

f

.De

DI9P

e O'I

-.0%

-. 08UZ

-eI2

-o I6

-~ 20

-r 2'I

-. 28

~ 32 IAE

~ 00050 . OOI 00 . ODI SO ~ 00200, 00250 . 00300, 00350, 00900 . OOIISO ~ OOSOO

ACCE AEAC'IOR PAESSUAE VESSEL -- UNIFOAII DAOP, IS FT. ELASTIC

FIGURE 22: Displacement (inches) of Node 61 at the Bottom of theVessel for a Uniform Drop versus Time (seconds)

160

FORC

Ieo

I 'lO

120

100ERCT ION

00

IHE

0 .00050 F 00100 .OOISO «00200 .00250 .00300 ~ 00350 F 00300 F 00'150 .OOSOO

RCCE RERCIOR PRESSURE VESSEL -- ONIFORN OROP. 15 F1. ELRSFIC

FIGURE 231 Reaction Force (lbs) at the Nozzle Support for a UniformDrop versus Time (seconds)

250

STRS

225

200

175

150

I25

TOO

75

0 SETN

50

25

0 .00050 . 00100 . OOISO . 00200 . 00250 .00300 ~ 00350 .00'tIOO .00%50

INE

.00500

'C(E

REACTOR PRESSURE VESSEL -- ONIFOBN DROPS IS FTe ELASTIC

FIGURE 24I Equivalent Stress (psi) in Element 60 at Node 35versus Time (seconds)

RG4 REACTOR PR SSURE VESSEL, OG OPgP, FROM 16 FT. (OCP-M3I

FIGURE 25: ANSYS Finite Element Model for Concentrated DropAnalysis — 3D Hidden Line Plot of Complete Model

RGltE REACTOR PRESSURE VESSEL. EOGE OROP FROR ]6 FT. (DCP-"3)

FIGURE 26: ANSYS Finite Element Model for Concentrated Drop Analysis—3D Hidden Line Plot of the Vessel Flange and Upper Shell

RGCE REACTOR PRESSURE VESSEL HOOEL, 30 OEG. SEGHENT lSCP-E-H3)

FIGURE 27: ANSYS Finite Element Model for Concentrated Drop Analysis-Side View of Complete Model

RG4E REACTOR PRESSURE YESSEL HOOEL. 30 OEG. SEGHENT tSCP-E"R3)

FIGURE 28: ANSYS Finite Element Model for Concentrated Drop Analysis-Front View of Complete Model

12

10

26

25

21

22 23

35

37

20

5556

57

6567

89

11190ll91

7576

92

77

11593

AGEE REACTOR PRESSURE VESSEL HOOEL, 30 OEG. SEGHENT (SCP-E-H3)

FIGURE 29: Element Numbers for the Flange and Upper Shell Portionsof the Concentrated Drop Model

e7

es

10

8889

75

7e

77

112

90

92

. 114

115 »e

120 117 119

121 118

AGEE REACTOR PRESSURE VESSEL MODEL. 30 OEG. SEGMENt lSCP-E-M31

FIGURE 30: Element Numbers for the Lower Half of the Upper ShellPortion of the Concentrated Drop Model

6

0 11 12 13

1'1 15 16 17

18 19 20 21

2'1 25

"~,.+ 5 /3Q 31

29 /32 33 3'9

/A y/ ~ 3e VO

;~i x / XM. 3 ~3~ S3

)( Nh„,W

ApproximateLocation ofSafety InjectionNozzle (Node 34)

!15 '96 'l7 t8RGCE REACTOR PRESSURE VESSEL HOOEL. 30 OEG. SEGHENT (SCP-E-H3)

FIGURE 31: Node Numbers for the Flange and Upper Shell Portions ofthe Concentrated Drop Model

h 8 ~ I' I

32 33 39

37

i<'i/'

93 99 99

, 39 50RG4E REAC70R PRESSURE VESSEL MOOEL, 30 DEG. SEGMENT (SCP-E-M3)

FlGURE 32: Node Numbers for the Lower Half of the Upper ShellPortion of the Concentrated Drop Model

122 123

12'1 125

126 127

128 129

RG4E REACTOR PRESSURE VESSEL HOOEL, 30 OEG. SEGMENT (SCP-E-H3)

FIGURE 33: Element Numbers for the Extended Intermediate Shell Portionof the Concentrated Drop Model

ll

f

r

4 't

HS 36 '%2 'HB

/N/XI. 9 SQ 51

52 53

5'%5

Se 57

Sa 5~ eO

*

et ea e3RGCE REACTOR PRESSURE VESSEL HODEL. 30 DEG. SEGHENT ISCP-E-H3)

FIGURE 34: Node Numbers for the Extended Intermediate Shell-Portion of the Concentrated Drop Model

28

27

29

30

31

32

38 90

SI

50 52

58 e0 e2

e8 7l 73

69 70 72

78

79 82

83 88

87

RGjtE REACTOR PRESSURE VESSEL HOOEL, 30 DEG. SEGHENT (SCP-E"R3)

FIGURE 35: Nozzle Element Numbers for the Concentrated Drop Model—Side View

r

~ 95

96

97

98

10

101

10

10

79 82 85 87

78 86

68

70

71

72

73

AGEE REACTOR PRESSuRE VESSEL HOGEL. 30 OEG. SEGMENT (SCP-E-H3)

FIGURE 36: Element Numbers for the Lower Half of the Nozzle - TopView, Concentrated Drop Model

~ *

I't 1 ~ ll

72; 81

IO

90 26

70 88

69 78 87 29

68 77 86

67 76 85 38

L IXX/79 96 33 92 36

AGEE REACTOR PAESSUAE VESSEL HOOEL. 30 OEG. SEGHENT (SCP-E-H3)

FIGURE 37: Nozzle Node Numbers for the Concentrated Drop Model—Side View

A 1

73 33 82 91 35

7'6 BB 92—36

75

ve 65

RGCE REACTOR PRESSURE VESSEL MODEL, 30 DEG. SEGMENT (SCP-E-M3)

FIGURE 38: Node Numbers for the Lower Half of the Nozzle - Top View,Concentrated Drop Model

FORC

'IO

20

-20

-'IO

60

-80

-IOO

-I20

-I'I0

.0020,00LI0,0060 .0080 .0100 .OI2D

INE

.OIVO .OI60 .OI80 .0200

ACUTE AEAC)OR PRESSURE VESSEL, EOCE OAOP FAON I6 FF. IDCP-NSI

FIGURE 39: Contact Force (lbs) between the Reactor Head and VesselFlange for a Concentrated Drop versus Time (seconds)

~ 20

OISP

- ~ 20

-. '%0

V

I V

-.eo

-. 80

1. 00

-I ~ 20

0 V

-I ~ 10

-I, 60

1. 80

. OO2O . OOOO . 00eO ..OOOO . OIOO . O12O . 01 ~0

IKE

.0180 .0180 .0200

RCtE REACTOR PRESSURE VESSEL. EOCE OROP FROH IE FT. IOCP-K31

FIGURE 40: Displacement (inches) of Nodes 10,11, 12 and 13 at the Topof the Reactor Vessel Flange for a Concentrated Drop versusTime (seconds)

OISP

.Ie

F08

-. 08

-. Ie

- ~ 2'I

-. 32

-s90

.98

-.se.0020 .0090 .0060 .0080 .OIOO .OI20

IIIE

.01'ID .OleD .0180 .0200

AGCE RERC'IOR PRESSURE VESSEL. EOCE DROP FROH l6 F'I. (DCP-P3)

FIGURE 41: Displacement (inches) of Node 61 at the Bottom of the Vesselfor a Concentrated Drop versus Time (seconds)

200

ENGT

180

160

1'40

120

100

BO

60

20

THE

0 .0020 .0090 .0060 .0080 .0100 .0!20 .01'90 .0160 ~ Ol80 .0200

AOCE AEACTOA PAESSUAE VESSEL, EOCE OAOP FAOkl 16 FT ~ (OCP-llew)

FIGURE 42: Kinetic Energy (inch-pounds) of the Reactor Head versusTime (seconds)

'h

lt

500

EkDT

E IH LB

(100

550

SOD

25D

200

150

10D

50

0 002D .00(IO 0060 .0080 ~ 0100 .0120 .DIED ~ 0160 ~ Ol SO .0200

RO<E REACTOR PRESSURE VESSEL, EOOE DROP FROII 16 FT. IOCP-IISI

IIIE

FIGURE 43: Strain Energy (~h-pounds) Absorbed by Element 4 versusTime (seconds)

~ I ~

I ~

~ I ~

I ~

I I ~

I ~

I I

~ ~ ~ ~ ~ ~ ~

~ ~

180

FORC

160

190

120

100

80

60

%0

20

ERCT Ott

-20 1 IIE

0 .OO2O .OOSO .OO6O .OOeO .OTOO .O12O .OTTO .O16O .O18O .O2OO

AG4E REACTOR PRESSURE VESSEL, EOGE OROP FROtl 16 FT. (OCP-A3)

FIGURE 45: Reaction Force (lbs) at. the Nozzle Support for aConcentrated Drop versus Time (seconds)

oOS

SIAN

. 0'I

e O'I

06 E PF

.03

F 08

.02

+02

F 01

06 E PB

oOI

.OD

INE

0 .0020 ~ 0090 .0060 .0080 .OIOO .OI20 .OI90 ~ OI60 .OI80 .0200

ROCE AERCIOR PRESSURE VESSEL, EOCE DROP FADE l6 FI. IOCP-NSI

FIGURE 46: Equivalent Strain at the Top (Outside Face) andBottom (Inside Face) at Element 106 for aConcentrated Drop versus Time (seconds)

17. 5

RATE

15.0

12+5

10.0

7.5

5.0

2a5

EPI-

-2. 5

-5.0

-7.5. 0020 . 00'IO . 0060 . OOBO . 0100 . 0120 . 0 I IIO

IHE

. 0160 . 0 I BO . 0200

AG4E RERC'IOR PRESSURE VESSEL. EOGE OROP PRON 16 FT. IOCP-R31

FIGURE 471 Equivalent Strain Rate (in/in/sec) at the Top (Outside .

Face) of Element 106 for a Concentrated Drop versus Time(seconds)

~ 16

0 1 SP

.08

-.08

~ 16

o 2'I

~ 32

-.40

-.'I8

-.S6

-.64.0020 .0040 .0060 .0080 .0100

I lIE

. 0120 . 0140 ~ 0160 ~ 0180 . 0200

ROTE RERCIOR PRESSURE VESSEL, EDGE DROP FAOII 16 FI. IOCP-1131

FIGURE 48: Vertical Displacement (inches) of Node 34, which is theapproximate location of the Safety Injection Nozzle,versus Time (seconds)

~ 18

DISP

.16

.12

.DR

.02

IHEo02

0 ~ 0020,00IIO ~ 0060 ~ 0080, 0100 ~ 0120 ~ 0 I IIO . 0160 ~ 0180 ~ 0200

RCCE RERCIOR PRESSURE VESSEL, EDGE DROP FADH'6 FT. IOCP-II3)

FIGURE 49: Radial Displacement (inches) of Node 34,which is the approximate location of theSafety Injection Nozzle, versus Time(seconds)

F00

D]SP

«00

.00

.00

-.OD

F00

-.00

-.00

-.00

.00.0020 ~ 0090 .0060 ~ 0080 .0100 .0120 .Ol'IO eOI60 ~ 0180 .0200

IIIE

ROTE RERCIOR PRESSURE VESSEL. EDGE DROP FROR 16 F I ~ IOCP-IISI

FIGURE 50: Rotation (radians) about the Circumferential Axis of Node 34,which is the approximate location of the Safety InjectionNozzle, versus Time (seconds)

%00

VEL

820

2%0

160

E40- EL

80

-80

-160

-BO

-820

JIVE-'lop

0 .0020 .opgp .0060 . popo .Oooo . Oi20 .Oiso .0>eo .Oiop .o2oo

RG4E REACTOR PRESSURE VESSEL, EDGE DROP FROH lb FT. (DCP-l',3)

FIGURE 51: Reactor Head Velocity (inches/second, Node 6)versus Time (seconds)

ATTACHMENT 3

Reactor Upper Internals Drop Analysis

MF3J

Technical Report No. 83084-2Revision 0

FINAL REPORT

FOR THE

UPPER INTERNALS DROP ANALYSIS FOR GINNA STATION

JOB NO. 83084

R. E. GINNA NUCLEAR POWER STATION

ROCHESTER GAS AND ELECTRIC CORPORATIONROCHESTER, NEW YORK

COPY NO.

f,, It.,

Prepareda Engine r

.) 3o g4

Approvedobject Manager

l go 5$

Approved CM:lf ~ 9~. ~ ~ o/8foc ester as " ectric

. g-Fr

Cygna Energy Services600 Atlantic Avenue

Boston, Massachusetts 02210

TABLE OF CONTENTS

Section

1. 0 Introduction

2.0 Upper Internals Description

3.0 Problem Description

4.0 Acceptance Criteria5.0 Analysis Methodology and Results

5.1 Energy at Impact5.2 Fuel Assembly Buckling5.3 Upper Support Plate Deformation5.4 Other Considerations

6.0 Conclusions

Pacae

2

58

1012

13

A endices

AB

ReferencesFigures 1 through 12

J ~

FINAL REPORT

U er Internals Dro Anal sis

1.0 INTRODUCTION

This report documents the results of Cygna's analysis of theupper internals dropping into the core barrel of the reactor atGinna during refueling operations. The report is intended toform the basis for a response to issues raised in NUREG-0612.

Three basic analyses were conducted. The first was an analysisto determine the total impact kinetic energy developed by theupper internals as it falls freely through water within thereactor cavity pool and reactor core barrel. The second was a

non-linear finite element plasticity analysis of the uppersupport plate of the upper internals performed to determine thestrain energy absorbed upon impact. The third was an elasticbuckling analysis of the fuel assemblies performed to determinethe additional strain energy absorbed when the fuel assembliesare impacted by the upper internals. The total strain energyabsorbed through plastic deformation of the upper support plateand elastic buckling of the fuel assemblies was compared to theimpact kinetic energy. The report concludes that a postulateddrop of the upper internals from a height of fifteen feet abovethe core is acceptable.

2.0 UPPER INTERNALS DESCRIPTION

The upper internals package at Ginna is a cylindrical structureconsisting of top and bottom plates separated by a series of ver-tical support columns (Figure 1) and the control rod guide tubes.The top plate, called the upper support plate, is four inchesthick and approximately 128 inches in diameter. The bottomplate, called the upper core plate, is one and one half inchesthick and approximately 108.5 inches in diameter. The plates arestructurally separated by 12 vertical support columns approxima-tely 115 inches in length. The entire upper internals package issupported by the outer edge of the upper support plate which sitson an "0" ring at the top of the core barrel flange (i.e., theupper internals "support ledge" ), (Figures 2 and 3).

3.0 PROBLEM DESCRIPTION

During refueling operations, the upper internals are removed fromthe core barrel using a special lifting device attached to the100 ton containment overhead crane. The lifting device has threesupport legs which-engage three threaded holes evenly spacedalong the outer periphery of the upper support plate. To insurevertical alignment during the lift, the special lifting devicealso has three vertical sleeve attachments that fit over threewidely spaced guide studs which are threaded into the reactorvessel flange (Figure 4). The guide studs project approximately15 feet above the top of the upper support plate.

At any time while the upper internals package is directly overthe reactor vessel, there is a possibility that, if dropped, theupper internals could impact the fuel assemblies in the reactor

core. However, once the lifting rig guide sleeves are above thetop of the three guide studs, the upper internals can be moved

out of concentric alignment with the reactor vessel core barrel(Figure 5). With the upper internals sufficiently out of con-centric alignment with the vessel, it is impossible for the upperinternals to impact the fuel assemblies. At a height of 15 feet,the guide sleeves are approximately one foot above the top of theguide studs. This is the maximum height for which concentricalignment of the upper internals and vessel must be maintainedand is therefore considered the maximum height from which theupper internals could still impact the fuel assemblies in thereactor (Figure 6).

The weight of the upper internals was determined from the maximum

load cell readings recorded when lifting the upper internalspackage during several recent refueling outages. It was conser-vatively assumed that, in addition to the weight of the upperinternals (52,100 pounds), the total load dropped onto the fuelassemblies included the weight of the lifting rig (12,000 pounds)and the crane hook block (8,500 pounds). Thus, the total dryweight used in the drop analysis was conservatively taken as72,600 pounds.

4.0 ACCEPTANCE CRITERIA

The final basis for the acceptability of the consequences of a

postulated upper internals drop is that neither the fuel rodsshall rupture nor shall the fuel be crushed. Under these con-ditions, no radioactive gases can be released. To prohibit fuelrod rupture and fuel crushing, the following conservativecriteria were established:

1. The strain in the Zircaloy-4 fuel rod shall remain in theelastic range after axial and lateral deformation. Thematerial maximum strain is limited to the yield strain of0.0047 inches/inch. This is well below the allowable strainused for design of 0.01 inches/inch.

2. The strain energy absorbed in the Zircaloy-4 guide tube shallbe limited to the energy absorbed prior to elastic buckling.Energy absorbed during inelastic deformation shall not beconsidered.

3. The strain energy absorbed by the upper support plate shallbe strain limited such that the equivalent strain in thestainless steel upper support plate of the upper internalsshall not exceed one half of the minimum ultimate strain of0.4 inches/inch. Thus, the actual equivalent strain in theupper support plate shall not exceed 0.2 inches/inch.

If the above criteria are satisfied and the energy absorbedduring impact is greater than the kinetic energy of the inter-nals at impact, the consequences of the load drop are acceptable.

5 .0 ANALYSIS METHODOLOGY AND RESULTS

Three principal evaluations were performed. First, an analysiswas made to determine the total kinetic energy developed by theupper internals falling freely through water (both in and out ofthe core barrel) and impacting the upper internals support ledge.Next, a non-linear finite element plasticity analysis of theupper support plate was performed to determine the strain energyabsorbed upon impact with the upper internals support ledge.

Finally, an elastic buckling analysis of the fuel assemblies was

done to determine the additional strain energy absorbed when thefuel assemblies are impacted by the downward displacement of theupper internals. The total strain energy absorbed throughplastic deformation of the upper support plate and elasticbuckling of the fuel assemblies was then compared to the kineticenergy of the internals at impact.

5.1 Ener at Im act

The upper internals package was assumed to fall through a liftedheight of 15 feet, which is approximately one foot above theheight at which the lifting rig guide sleeves are decoupled fromthe three guide studs. At this point, the bottom of the upperinternals is approximately five feet above the vessel flange(Figure 5). The analysis assumed the upper internals wouldinitially fall through five feet of water above the vessel, fit

I

perfectly into the vessel core barrel (with only 1/4 inch clearspace all around), and fall through ten feet of water in the corebarrel before impacting the support ledge. It was conservativelyassumed that no binding or wabble would occur along either theguide studs or the core barrel that would tend to reduce theimpact kinetic energy (or possibly even prevent impact).

To determine the impact velocity of the upper internals at theinstant -it strikes the upper internals support ledge, thecalculation of the resisting (drag) forces acting on the upperinternals as it falls freely in water is required. The basicexpression for the resisting forces has the form

where f is the mass density of water, V is the velocity of fall,A is a projected area on a plane perpendicular to the directionof travel, and C is the drag coefficient which depends on theshape of the falling object and confining conditions. Dragcoefficients were derived for three separate regions: above thevessel, in the core barrel at the reactor coolant loop outletnozzle openings, and in the core barrel above and below thenozzle openings (Figure 7).

To determine the resisting force acting during the free fallabove the vessel, only the gross area of the upper support platewas used in conjunction with a drag coefficient for a solid cir-cular disc of l.0. The use of this drag coefficient togetherwith the gross area of onl'y the upper support plate is reasonablesince the reduction in drag due to the openings in the uppersupport plate is more than compensated by neglecting the drag onthe net area of the upper core plate.

The resisting force for free fall of the upper internals in thecore barrel was derived from first principles (i.e., theequations of motion, continuity and momentum). In the deriva-tion, the only portion of the upper internals on which resistingforces were considered to act was the upper core plate.Resisting forces acting on the upper support plate, supportcolumns and control rod guide tubes were neglected. Consideringthe upper core plate to be a flat circular ring falling freelywithin a fluid filled cylinder, the expression for the resistingforce becomes

I

2.

where V and g are defined as before, A

barrel opening andis the area of the core

is the "drag coefficient". Within this expression for the dragcoefficient, Ao is the area of all openings through which fluidcan pass from one side of the upper core plate to the other, and

Cd is the discharge coefficient which accounts for the effects ofthe vena contracta. The discharge coefficient generally rangesbetween 0.5 and 1.0. A value of 0.8 is reasonable for the uppercore plate and was used in the analysis.

When the internals enters the core barrel, the net flow area ofthe upper core plate for the region of the core barrel above theoutlet nozzles results in a computed drag coefficient of 5.4.As the upper core plate falls past the openings for the twooutlet nozzles, the net flow area is increased and the dragcoefficient becomes 2.1. The value 2.1 is conservative since itassumes that for the entire nozzle depth water can flow around theupper core plate through an area equal to one half of the nozzlesopen area. This is the maximum area and occurs only when theupper core plate is at the midheight of the nozzle opening. Once

the upper core plate reaches the bottom of the opening, a dragcoefficient of 5.4 is again used until impact.

Equilibrium of the free body diagram of the falling internalsresults in a second order non-linear differential equation. Thisequation was solved for velocity as a function of displacement.Assuming the velocity to be zero at the time of the drop, thevelocity was calculated at the beginning and end of each of thefour regions through which the internals must fall, as well as atseveral intermediate points. A plot of the internals velocityprofile during free fall is shown in Figure 8. The final velocityof the upper internals at the instant of impact with the fuelassemblies is 15.3 feet per second. This velocity results in a

total kinetic energy at impact of 263,000 foot-pounds.

5.2 Fuel Assembl Bucklin

Due to the small gap which initially exists between the upperinternals and the fuel assemblies and the fact that the uppersupport plate must deform by several inches after impact in orderto absorb 263,000 foot-pounds of energy, it was apparent thatsome of the impact kinetic energy must be absorbed by the fuelassemblies themselves. Thus, the analysis of the fuel assembliesprovided two important results: (1) the amount of elastic strainenergy that could. be absorbed by the fuel assemblies and (2) theamount of vertical displacement the fuel assemblies wouldexperience while absorbing this energy. The magnitude of thisdisplacement, together with the known gaps between the upperinternals and fuel assembly nozzles and between the nozzles and

fuel rods, determined the permissible vertical displacement ofthe upper internals. Knowing this permissible vertical displace-ment, the total strain energy absorbed by the upper support plateof the upper internals could be determined.

The actual analysis of the fuel assemblies conservatively con-sidered only the energy absorbed during elastic buckling. The

analysis assumed that the guide tubes buckled at their narrowestsection between the two lowest grid spacers and that the fuelrods buckled between all nine grid spacers which tie the guidetubes and the fuel rods together. Since the distance between thegrid spacers was somewhat variable, an average length was assumed

in the buckling analysis. An end restraint factor of 3.4 (4.0 isfixed and 1.0 pinned) was used for the guide tubes, while a

factor of 1.08 was used for the fuel rods. Both the guide tubeand fuel rod cladding material is Zircaloy-4.

The energy absorbed by the 16 guide tubes in axial deformationprior to elastic buckling was calculated for each assembly.

J

(The energy absorbing capability of the instrument tube has been

conservatively neglected in this analysis.) Based on the struc-ture of the fuel assemblies (Figure 9), the guide tubes must

buckle prior to the fuel rods because they attach directly to thefuel assembly nozzles, whereas a total gap of 1.565 inches existsbetween the nozzles and the fuel rods. After the guide tubecritical buckling load was reached, no further energy was assumed

to be absorbed by the guide tubes. Downward displacement stillcontinued until the 179 fuel rods per assembly were loaded.Again, the fuel rods were assumed to deform only to the elastic

Ilimit. The lateral deformation due to an applied axial load was

determined from an initial fuel rod bow of 0.001 of the span

length. The resulting strain energy from axial and bendingdeformation due to initial curvature was calculated. The com-

bined strain energy from the elastic buckling of the guide tubesand the fuel rods for each assembly was determined.

The elastic buckling of the 16 guide tubes in each fuel assemblyresults in 163 foot-pounds of absorbed strain energy. As the179 fuel rods per assembly are loaded, a total of 2,130 foot-pounds of strain energy is absorbed prior to reaching theelastic limit; Therefore, for each fuel assembly a total of2,293 foot-pounds of strain energy is absorbed during elasticdeformation. The total strain energy for all 121 fuel assembliesin the core (Figure 10) was thus determined to be 277,000 foot-pounds.

The analytic determination of the energy absorbed by the fuelassemblies is conservative since'o credit has been taken forenergy absorbed in the inelastic range. The yield strain valueof 0.0047 inches per inch for Zircaloy-4 material is well belowthe minimum allowable strain of 0.01 inches per inch used in the

design of the fuel assemblies. If the material were allowed togo into the plastic range, a significant amount of additionalstrain energy would be absorbed. Furthermore, additional defor-mation in the fuel assemblies due to non-linear effects wouldresult in significant additional energy being absorbed by theupper internals support plate, since the internals could deflectdownward through an additional finite distance.

From the displacement of each fuel assembly due to axial short-ening, the total downward displacement of the upper internals was

computed. This displacement in turn determined the final strainenergy absorbed by the upper support plate.

5.3 U er Su ort Plate Deformation

The upper support plate was analyzed using the elastic-plasticfinite elements available within the ANSYS program, Version 4.1B.Due to symmetry considerations, only a 45'egment of the uppersupport plate was modeled. The mass of the attached structuresbelow the upper support plate (support columns, upper core plate,etc.) as well as the mass of the load block and lifting rig wereconservatively distributed over the inner regions of the platedirectly above the open core barrel. The basic geometry of theinitial finite element mesh was determined from several testproblems 'for which the results could be compared to known exactsolutions. A fine finite element mesh was not required sinceonly displacement and strain energy calculations were required.Thus, an accurate yet cost effective mathematical model was

developed. An elastic static analysis was performed to insureproper model behavior.

With the results from the elastic analysis, the model was revisedto account for plasticity. Those elements identified as poten-

-10-

v

tially capable of developing plastic strains were modified fromelastic quadrilateral plate bending elements to triangularplasticity plate bending elements as required by the ANSYS

program (Figure ll). Static plastic analyses were then performedusing a series of applied accelerations ranging from O.lg to32g's. In this way, the maximum displacement was determined foreach value of acceleration. The total strain energy absorbed bythe upper support plate when then found by integrating thestresses and strains over the volume of each finite element.This provided a relationship between the maximum displacement ofthe plate (at any "g" level) and the total strain energyabsorbed.

The energy absorbed by the upper internals package is directlydependent on the deformation through which it moves. This defor-mation is equal to the distance between the bottom of the uppercore plate and the top of the fuel assemblies at the point wherethe elastic limit of the fuel rods has been reached. Quantita-tively, this distance is the sum of a 0.3 inch gap between the,upper core plate and the fuel assemblies under normal conditions,a 1.565 inch gap between the top and bottom fuel assembly nozzlesand the tips of the fuel rods, and a 0.549 inch axial shorteningof the fuel rods during elastic buckling. The total deformationis 2.41 inches. As shown in the graph of Figure 12, this defor-mation level corresponds to 107,000 foot-pounds of strain energyabsorbed by the upper internals support plate. The correspondingmaximum equivalent strain in the upper support plate at thisdeformation level is approximately 0.014 inches/inch. This iswell below the acceptance value of 0.2 inches/inch.

-11-

5.4 Other Considerations

In addition to the conservative assumptions already incorporatedin the analysis, three other important conservative assumptionshave been made in this evaluation. First, the upper internalsare free to fall without binding or wabbling. Clearly any smalllateral movement or twisting of the upper internals might be suf-ficient to prohibit the internals from falling entirely back intothe core barrel, since only a one quarter inch clearance existsbetween the core barrel and the upper core plate. As a minimum,

some energy will be dissipated each time the lifting rig sleevesslide against the guide studs, or the upper core plate scrapesthe core barrel.

Second, the mass of the 12,000 pound lifting rig is assumed to be

distributed over and carried by the upper support plate in thearea directly over the core, when, in fact, the lifting rig issupported by three relatively stiff leg assemblies and a supportring located directly above the support ledge of the core barrelflange. The kinetic energy due to the lifting rig could actuallybe applied directly to the core barrel flange, rather than to theupper internals and fuel assemblies as has been assumed in theanalysis.

Third, the kinetic energy loss which occurs during the transferof momentum between the upper internals and the fuel assemblieshas been neglected. This energy loss is probably on the order of60% of the total impact kinetic energy, but since it is generallydifficult to quantify, it has conservatively been neglected.

-12-

6.0 CONCLUSIONS

The results of the above analyses have conservatively estimatedthe total energy capable of being absorbed during an impact bythe upper internals as 384,000 foot-pounds. This consists of277,000 foot-pounds absorbed by the fuel assemblies duringelastic deformation and 107,000 foot-pounds absorbed by the uppersupport plate of the upper internals while deforming compatiblywith the fuel assemblies. The lower bound estimate of 384,000foot-pounds for the energy absorbed during impact is well inexcess of the 263,000 foot-pounds of kinetic energy possessed bythe upper internals at the time of impact. Since all of theabsorbed energy has been calculated from deformations which do

not exceed the elastic limit of the fuel rods, it is concludedthat a postulated drop of the upper internals back into thevessel would neither rupture the fuel cladding nor crush the fuelin the fuel rods.

Thus, a postulated drop of the upper internals from a height of15 feet will be acceptable. At this height, the bottom of theinternals is approximately five feet above the top of the corebarrel, and the lifting rig guide sleeves are above the top ofthe three guide studs Qy approximately one foot. It is very dif-ficult to conceive of the upper internals falling freely from anyhigher elevation without any interference, since the smallestlateral movement will preclude a unincumbered fall back into thecore barrel.

-13-

APPENDIX A

References

REFERENCES

2.

3.

4 ~

5.

6.

7.

8.

9.

10.

Fluid Mechanics, by V. L. Streeter, Third Edition,McGraw-Hill Book Co., 1962.

Ordinary Differential Equations, by W. Kaplan,Addison-Wesley Co., 1958.

Theory of Elastic Stability, by S. P. Timoshenko and J. M.Gere, Second Edition, McGraw-Hill Book Co., 1961.

ANSYS Engineering Analysis Systems User's Manual, Revision4.1, Swanson Analysis Systems Inc., Houston, PA.

ANSYS Engineering Analysis Systems Theoretical Manual,November 1977, Swanson Analysis Systems, Inc., Houston, PA.

Kent's Mechanical Engineers Handbook, J. K. Salisbury,Twelfth Edition, John Wiley and Sons Inc., 1950.

Letter: M. B. Fitzsimmons (RG&E) to J. D. McWilliam(Cygna), 8/22/83.

Letter: M. B. Fitzsimmons (RG&E) to J. D. McWilliam(Cygna), 10/17/83.

Letter: M. B. Fitzsimmons (RG&E) to J. D. McWilliam(Cygna), 12/13/83.

Letter: M. B. Fitzsimmons (RG&E) to J. D. McWilliam(Cygna), 1/16/84.

11. Gilbert Associates .Inc., Drawing D-521-064.

12. Westinghouse Electric Corp. Drawings, 541-F-205, 541-F-321,541 F 581'84 J 650'84 J 652'84 J 673g 684 J 732/684-J 783~ 684 J 784~ 684 J 852~ 684 J 988, 685 J 464.

APPENDIX B

Figures 1 through l2

LIST OF FIGURES

1. Upper Internals

2. Upper Internals within the Reactor Vessel

3. Upper Internals Support Ledge

4. Upper Internals Being Raised from the Reactor Vessel

5. Upper Internals Decoupled from Guide Studs

6. Upper Internals Drop into the Reactor Vessel

7. Regions within Upper Internals Drop Path

8. Upper Internals Velocity During Free Fall

9. Fuel Assembly

10. Plan View of Core with 121 Fuel Assemblies

ll. ANSYS Finite Element Model of Upper Support Plate

12. Energy Absorbed by Upper Support Plate

127.855

UPPER SUPPORTPLATE

SUPPORTCOLUMN

0

CO(0Al

108.585

UPPER COREPLATE

FIGURE 1: UPPER INTERNALS

Note: Control Rod Guide Tubes not Shown for Clarity

'~ ~

GUIDESTUDS~O

I

LIFTINGRIG

a

l1

a

'a'atl

optic

aaaaa a

ll

llll

II

ll

Il

DETAIL A

FIGURE 2. UPPER INTERNALS WITHIN THE REACTOR VESSEL

GUIDE STUD

UPPER SUPPORT. PLATE

0 RING

VESSELFLANGE

CORE BARREL

DETAIL A

FIGURE 3. UPPER INTERNALS SUPPORT LEDGE

\

GUIDESTUDS~

4—LIFTINGRIG

0I

Ul

00.00a'0

II

ll

p

0000Ijoaoo,a,00~0

FIGURE 4. UPPER INTERNALS BEING RAISED FROM THE REACTOR VESSEL

4—LIFTINGRIG

ll

tj

Pp

II

0 0

p'a000

0

tj

ll

Ilp

[]

pp

za0IUl

tl0I

IA

FIGURE 5. UPPER INTERNALS DECOUPLED FROM GUIDE STUDS

~ LIFTIN GRIG

Il

0

0

0

0',00

00(

Il

Il

Il

0

0

0

0

0

00

0

0

FIGURE 6. UPPER INTERNALS DROP INTO THE REACTOR VESSEL

Reactor CavityPool ~LIFTING

RIG

Core BarrelAbove Nozzles

Core BarrelAt Nozzles

Core BarrelBelow Nozzles

0

0

'0ooDp00

0

o'oo'oo'oo'o

pl

FIGURE 7. REGIONS WITHIN UPPER INTERNALS DROP PATH

r

20

'18

IMPACTVELOCITY

16

14

—15.3 ft/sec

~ 12

~~ 10

8u0

6

00 1 DISTANCE (FT.)

Reactor CavitPool

Core BarrelAbove Nozzles

Core Barrel Core BarreAt Nozzles elow Noz

FIGURE 8. UPPER INTERNALS VELOCITY DURING FREE FALL

HOLD DOWN SPRING

TOP NOZZLE

CONTROL ROD

)()(l illill (l(l( ~l(I'UELROD

SPACERGRID

'l(l(lI()(l„

((i( l()(I ()(,)()()(" ~

g)l);)(

GUIDE TUBE

'ill( il()

'((((((()((()(

GRID SPRING ~ ((,g~ ((g((

„l()(l')(l

()(

BOTTOM NOZZLE

FIGURE 9. FUEL ASSEMBLY

CORE BAFFLE CORE BARREL

THE Rt,'~AL SHIELD

REACTOR VESSEL F UEL AS S.'li'.BLY

I I 2. 50 O. D

I ll. 50 I. D.

I22. 50 O.D.132. 00 I. D.

FIGURE 10. PLAN VIEW OF CORE WITH 121 FUEL ASSEMBLIES

8 8

8 8

88

8990

71 7

7 777

7 79

80 8

5 5

5 5 60

62 6

6es 6

6 6870

53

2 2 262

30

17

18

131

2 3

AGEE UPPER INTERNALS SUPPORT PLATE NCDEL I'15 OEG. SEGMENT)

FIGURE 11. ANSYS FINITE ELEMENT MODEL OF UPPER SUPPORT PLATE

/

5

~ - ~

"+ ~ ~

1,000,000454 4 5 ~ iiil

ri"i5 ~,I

ti Ii I ~

54

i !j j I ,I !i~!II:-.'S'W

~

1 ~

5 5

5

5-'j5-5-5-.4t

ZU ''t„} Iil,

5;

100,000I- ~

'I451

'I51

I'5

S

jc at~ »

~ }

.1:it!c.;

~ r 4

It 5

Wir)n

. 9 55}I'5 5

.tran J.t,Qi"- 3

DD':—~

~ 4 .5 ...t

10,000I 4

'5

~ r

'I'I I 55-"~ 5

5 5

~ ~ - ~ L

I, ~

"l~ ~

~ ~ 5 ~—

55 4

5'

+" Ii

.5

I ~ It'I I

'54: W ,I!}

.='- ..'i..'. 45»

t t ~

triit="I !~-t 4.

'-,-'.}5"!. }

55~

\ ~

4

1,000III 1'4 ~

:: 'cii

=':i-5

.ggcci

5 LL.'4 5—

""J

1'4'Ic

lt..:t"~}}t8:

-5'

~

j

~- ')!~i

= -tt ~ ~ 4

ii

5 ..

5

'-5

100fn

100.0Cl CO ir lO lh N ttl 55!

10.0MAXIMUMDISPLACEMENT (IN)

~ 4-4-

mar m u} w n cv

1.0 0.1

FIGURE 12. ENERGY ABSORBED BY UPPER SUPPORT PLATE

ATTACHMENT 4

Load/Impact. Matrix She'ebs

MF3J 19

LOCATION-

CRANE;" '9 Ton"Screenhouse.

"Screenhouse"East of"Column""5..

It'1PACT::

AREA

Service Water pump Motors and Associated Electrical Bus

LOADS

S. W. Pumps

S. W. Pump Motor

'. ELEVATION

Floor Eley.253<-6"

Floor Eley.253-6"

EQUIPNENT EVALUATED

S. W. System

:S. W. System

.:.. HAZARD ELININATION CATEGORY

See Discussion 2.4-2(b)

See Discussion 2.4-2(b)

Fire Pump Motors Floor Eley.253'-6"

:S. N. System See Discussion 2.4-2(b)

Fire Pump

Fire Pump Engine

Floor Eley.253r 6

Flqor Eley.253)-6

:S, W. System

:S. W'. System

See Discussion 2.4-.2(b)

See Discussion 2.4-2(b)

S. W. CheckValves

Floor Eley.253'-6"

:S. W. System See Discussion 2.4-. 2(b)

S. W. IsolationValves

Floor Eley.253'-. 6"

N. System See Discussion 2,4 2(b)

0

LOCATION..- .

CgtIE;.. 10 Ton"Jib in =contai t.

'ver'rezzurizer" Cavity.

If'1PACT:-

AREAPressurizer and Associated Piping/Instrumentation

LOADS .ELEVATION EQUIPMENT EVALUATED '-" HAZARD ELININATION CATEGORY

Hatch Covers Floor Elevatio288'-4"

:.Pressurizer:.Piping/instrumentation

See Response to NRC Section2-;4-2 in report Section 3.0

LOCATIOr<.

CpE ',". "2"Ton"Jib'n"Contai

"'At"Personnel.-Hatch"in"Containmerrt/

O.IMPACT:

AREA

LOADS . ---.". .-ELEVATION EtIUIPNENT -EVALUATED'

Operating Floor og Concrete and- Grating~ ~ ~

P

HAZARD-ELININATION CATEGORY

MiscellaneousEquipment

Floor Elevatio278 1-4 'I

(Maximum LoadHeight 11'-0")

.:B pccumulatoz*

.:Line te Loop 5

Zan Cooler D-and its servicewater lines* .-

h

See Discussion 2,4-2(b)

See Discussion 2. 4-2 (b)

*NOTE: The Sa ety-Related equipment listed i2 gloo levels below the crane and asbe dam ged.

locatedumed to

LOCATION.." ..

CPJPE;"'" nuxil'i'ary'uilding'ment Munorail'

" Basement; Auxiliary"Bui'lding;"Elev;"'235'-"8"'

Ir<PACT:

AREA Basemenh.,Floor/Equipment ..in..SM~..Basement

LOADS ." ELEVATION EQUIPMENT EVALUATED'"" .HAZARD ELIMki'i~ATION CATEGORY

RHR Pumps

RCDT Pumps

Removable FloorsSections

Floor Elev.

235'-8"RHR Pumps

RHR Piping

ee Discussion 2. 4-2 (b)

LOCATION . '..CRANE;" "3*Ton"Jib.

"'"Equ'ipme'n't"Ha't'ch/Crah'e"Bag'

IMPACT:

AREA

Operating Floor of Concrete andCrane Bay Opening/Basement Floor(Piping..Around..Crane..Bay).

LOADS

Misc. Equipment-,

.ELEyATION

Floor Elev.274'-6" See Discussion 2.4-2 (b)Mechanical Equipment

'A:accumulator line to B loopRHR letdown from A loopLPSl to 852ASX to B cold leg-charging line valve 296 and as@fqn coolers B and C'and associhB -.RCP seal water injectionElectrical Circuits

ociated pipingted service water

EggIPgENT EyALtjATED. — ",-.- .HAZARD .ELININATION CATEGORY

NOTE:

SG A narrow range level 461, 4SG A wide range level 460SG B narrow range level 471, 4LP'SI valve 852A, circuits C729LPSI valve 852B, circuits C109fan cooler B circuit L219fan cooler C circuit L228pressurizer pressure PT 430, Ppressurizer level LT427, LT428CVCS valves 294 circuit, R536 c

392A circuit R550200A circuit R515200B circuit R508-202 circuit R501 1296 circuit R543 p

Some safety-related equis located 2 floor levecrane and assumed to be

3:C731

,'731

431

arginghprgingetdownetdowntdown.essurizer spraypment listeds:.below thedamaged.

: LOCATION.........

h h'LCR/lt<E'0/5 Tori"Ai~xi.liary Buil . OverHead

- "..Auxiliary.Building-

II"/PACT:

AREA

LOADS

Operating Floor/Crane Bay

ELEVATION EQU I Pi'1EQT EyAI UATEB.---,— -

HAZARD -ELAN IRISATION CATEGORY

RHR Pumps

Component CoolinPumps

RCDT Pumps

RemovableFloor'ections

FloorElevations

271'-0" and278'-4"

RHR

C'omponent Cooling

See Discussion in -response toNRC 2.2-4 located in reportsection 3. Also see discussion2. 4-2 (b) .

Vibration Cleane

New FuelAssemblies

New .Fuel Casks

Steam GeneratorTools

Transfer Canal:Gate

Lead Basket

Shipping Cask'-

CARPE" 100/20" Ton Containmvenhead ."

LOCATION... "Containment,--.

IV!PACT:

AREA

Vessel/Cavity Floor/Operating Deck

LOADS

RPV Head onVessel

RPV Head onCavity Floor andHead Stand

"ELEVATION

l6'iftedabove vesselflange

Lifted 42'ax.above cavity.8'bove headstand

EQUIPMENT EVALUATED.

:Reactor Pessel

Mechanical Equipment

:Reactor coolant system:Accumulators.:RHR:LPSZ:SX:.Containment fan coolers:.CVCS charging letdownSG blowdown:FeedwaterMainsteam:RCP seal in)ection:Pressurizer spray- and: auxiliary spray

:.Electrical Circuits:SG leVel wide & narrow range:Pressurizer pressure & leVel:LPSl 852 valves:Fan coolers.Charging valVes 294 .& 392ALetdown valves 200A, 200$ ,

: 202, 427pressurizer PORVs 430, 43 C'Pressurizer spray valves

296, 431A, 431B

".:"..HAZARD EL'ININATION CATEGORY

See Response to NRC Section2.:.3-4 Located in Report Section3'. 0.

See Response to NRC Section2;4 g located in Report Section3;0.

IV

1

CQNE,'00/20"Ton 'Containm verhead'" """

"Containment

II")PACT

AREAVessel/Cavity Floor/Operating Deck

LOADS .—.—..—:..." ELEVATIO EQUI PHENT EVALUATED"

HAZARD EL"ININATION CATEGORY

Reactor UpperInternals onVessel

Reactor UpperInternals onNorth CavityFloor andStorage Stand;

Lifted15'bovethe core

rifted 15'.above cavityfloor,

Vessel Core

..REiR 'Outlet'rom "A" Hot Let

LPSI 852A Pipe to VesselUpper Plenum.

See Response to NRC Section '

2;3-4 located" in Report Section3:0.

~~'y'

V

W l,~f~J )

V

8(