ML20083M310

Control of Heavy Loads - Phase Ii,Arkansas Nuclear One, Units 1 & 2, Draft Technical Evaluation Rept

ML20083M310
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/02/1984
From:	Ahmed N FRANKLIN INSTITUTE
To:	Singh A NRC
Shared Package
ML20083M312	List: ML20083M310
References
CON-NRC-03-81-130, CON-NRC-3-81-130, REF-GTECI-A-36, REF-GTECI-SF, TASK-A-36, TASK-OR TAC-52206, TER-C5506-442-4, TER-C5506-442-443-DR, NUDOCS 8404170572
Download: ML20083M310 (30)

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TECHNICAL EVALUATION REPORT CONTROL OF HEAVY LOADS - PHASE II ARKANSAS POWER AND LIGHT COMPANY ARKANSAS NUCLEAR ONE UNITS 1 AND 2 NRC DOCKET NO. 50-313, 50-368 FRC PROJECT C5506 NRC TAC NO. 52206, 52207 FRC ASSIGNMENT 19 N RC CONTRACT NO. NRC-03-81 130 FRC TASKS 442, 443 Prepared by Franklin Research Center Author: N. Ahmed 20th and Race Streets Philadelphia, PA 19103 FRC Group Leader: I. H. Sargent Prepared for Lead NRC Engineer: A. Singh Nuclear Regulatory Commission  ;

Washington, D.C. 20555 April 2, 1984 This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty. expressed or implied. or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, appa-ratus, product or process disclosed in this report. or represents that its use by sucn third party would not infnnge privately owned rights.

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TER-C5506-442/443 f CONTENTS e

Section Title Page 1 INTRoooCTroN . . . . . . . . . . . . . 1  !

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1.1 Purpose . . . . . . . . . . . . . 1 1.2 Generic Background. . . . . . . . . . . 1

• 1.3 Plant-Specific Background . . . . . . . . . 2  !

2 EVALUATION . . . . . . . . . . . . . . 3  !

2.1 Evaluation Criteria . . . . . . . .. . . 3 a

2.2 Overhead Handling Systems in Spent Fuel Pool Area . . . . . . . . . . 4 2.3 overhead Handling Systems in Reactor vessel Area . . . . . . . . . . 8 I

- 2.4 Overhead Handling Systems in Areas Containing Safe Shutdown Equipment . . . . . . . . . 12 3 CONCLUSIONS . . . . . . . . . . . . . 16 h

- 3 .1 Information Isssues . . . . . . . . . .- 16 3.2 Approach Issues . . . . . . . . . '. . . 17 a

4 REFERENCES . . . . . .. . . . . . . . . 18 APPENDIX A - LOAD / IMPACT AREA MATRIX 4

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TER-C5506-442/443 FOREWORD This Technical Evaluation Report was prepared by Franklin Research Center under a contract with the U.S. Nuclear Regulatory Commission (Office of Nuclear Reactor Regulation, Division of Operating Reactors) for technical assistance in support of NRC operating reactor licensing actions. The technical evaluation was conducted in accordance with criteria established by the NRC.

Mr. I. H. Sargent contributed to the technical preparation of this report through a subcontract with WESTEC Services, Inc.

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1. INTRODUCTION ,

i 1.1 PURPOSE This technical evaluation report documents a review of load handling equipment operated in the vicinity of spent fuel and equipment employed for i

ro:ctor shutdown and fuel element decay heat removal at Arkansas Nuclear One Units 1 and 2. This review constitutes the second phase of a two-phase review in::tituted to resolve a generic issue pertaining to the safe handling of heavy loads at nuclear power plants.

1.2 GENERIC BACKGROUND Generic Technical Activity Task A-36 was established by the Nuclear Regulatory Commission (NRC) staff to systematically examine staff licensing criteria and the adequacy of measures in effect at operating nuclear power plants to ensure the safe handling of heavy loads and to recommend necessary changes in these measures. This activity was initiated by a letter issued by th3 NRC staff on May 17, 1978 [1] to all power reactor licensees, requesting  ;

information concerning the control of heavy loads near spent fuel. i l

The results of Task A-36 were reported in NUREG-0612 [2]. The staff j c ncluded from this evaluation that existing measures to control the handling l.

of heavy loads at operating plants provide protection from certain potential  !

problems but do not adequately cover the major causes of load handling  ;

cccidents and should be upgraded.

To upgrade measures for the control' of heavy loads, the staff developed a tcries of guidelines to implement,a two-part objective. The first part of the objective, to be achieved through the implementation of a set of general guidelines expressed in NUREG-0612, Section 5.1.1, was to ensure that all load '

handling systems. at nuclear power plants have been designed and are operated so that 'their probability of failure is appropriately small for the critical- j-tecks -in which _ they are employed. The results of the reviews associated with this part of the staff's overall objective were provided in a series of technical evaluation reports identified as Phase I reports. The~second part i-3 Frankun Research Center

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TER-C5506-442/443 of the staff's objective, and the subject of this report, was to be achieved through guidelines expressed in NUREG-0612, Sections 5.1.2 through 5.1.5. The purpose of these guidelines was to ensure that in the case of specific load handling systems used in areas where their failure might result in significant consequences, either (1) features have been provided, in addition to those required for all load handling systems, to make the potential for a damaging lead drop extremely small or (2) conservative evaluations of load handling tecidents indicate that the potential consequences of a load drop are cceeptably small.

1.3 PLANT-SPECIFIC BACKGROUND q On December 22, 1980, the NRC issued a letter (3] to Arkansas Power and Light Company (AP&L) , the Licensee for Arkansas Nuclear One Units 1 and 2, 4

coquesting the review of provisions for handling and control of heavy loads, the evaluation of these provisions with respect to the guidelines of NUREG-0612, and the provision of certain additional information to be used for En independent determination of conformance to these guidelines. The results of this independent evaluation with respect to general load handling equipment

- cnd procedures (Phase I) were provided on December 21, 1981 [4]. On December 22, 1982, AP&L provided a final Phase II report (5] concerning conformance to

. staff guidelines for specific load handling systems operated in areas where a load drop might result in significant consequences. That report provided the basis for this technical report.

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TER-CS506-442/443 2

2. EVALUATION This section presents an evaluation of critical load handling areas at Arkansas Nuclear One Units 1 and 2. Separate subsections are provided to identify the criteria used in this evaluation and each of the plant areas considered. For each such area, relevant load handling systems are identified, Licensee-provided information related to the evaluation criteria or proposed alternatives is summarized and evaluated, and a conclusion as to the extent of compliance, including recommended additional action or 3

requirements for additional information as appropriate, is provided.

2.1 EVALUATION CRITERIA i The objective of this review was to determine if plant arrangements and load handling equipment design were such that either the likelihood of a load handling accident that could damage spent fuel or . equipment used in reactor

. shutdown or fuel element decay heat removal is extremely small or that the consequences of such damage, should it occur, will be acceptable. Guidance contained in NUREG-0612, Sections 5.1.2, 5.1.3, and 5.1.5 ,(for pressurized Lwater' reactors) and in 5.1.4 and 5.1.5 (for boiling water reactors) forms the basis for the conclusions reached in this section and is briefly summarized as l follows.

For a determination that the likelihood of damage is extremely small:

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I o' The design 'of the load handl.ing system (i.e. , crane or hoist and underhook-lifting devices)'is consistent with, or equivalent to, th'e

NRC ' staff criteria for single-failure-proof cranes identified in

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NUREG-0554 [6), or

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' o The plant physical arrangement is such that a- crane operated in the vicinity of spent fuel .or safety-related equipment is prevented from travelling to a position from which a load-drop can be expected to -

damage-such equipment.

[ For a determination that the potential consequences of damage following a load drop will be acceptable:

I o In the case of potential damage'to spent fuel,. calculations have been provided to demonstrate that potential radiological doses at'the site

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. i TER-C5506-442/443 i boundary will not exceed 25% of the limits specified in 10CFR100 and that the post-accident configuration of the fuel will not result in a K,gg larger than 0.95.

o In the case of damage to the reactor vessel or spent fuel pool, it can be demonstrated that this damage will be limited to the extent that the fuel will not become uncovered.

o In the case of damage to equipment or components employed for reactor  !

shutdown or fuel element decay heat removal, it can be demonstrated that the safety-related function of the affected system will not be lost. '

2.2 OVERHEAD HANDLING SYSTEMS IN SPENT FUEL POOL AREA 2.2.1 Identification of Overhead Handling Systems

a. Summary of Licensee Statements and Conclusions The Licensee identified [5] the following load handling systems as being subject. to Phase II criteria of NUREG-0612:

o - fuel handling cranet (L3) o auxiliary fuel handling crane o new fuel handling crane ' (2L35) .

The fuel handling crane (Equipment No. L3) was manufactured by PEH Harnischfeger and is a 100-ton electric overhead traveling bridge crane with a 10-ton auxiliary hoist. A 2-ton auxiliary hoist is suspended, monorail-fashion, from a 12-inch I-beam welded between the end trucks at the south end of the fuel handling crane (L3) and.is designated as the Unit 1 auxiliary fuel handling crane. The Unit 2 new fuel handling crane (Equipment No. 2L35) was

manufactured by Heco-Pacific Manufacturing Company and is a 4-ton capacity, top-riding, single-girder crane consisting of a bridge, monorail, trolley, and )

hoist.

The weight of a heavy load is noted by the Licensee to be any load weighing more than 2000 lb.

l None of the - identified cranes are in' complete compliance with NUREG-0612, Section 5.1.6 (single-failure-proof) or in partial compliance supplemented by cuitable alternative'or additional design features. l l

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b. Evaluation and Conclusion The Licensee's identification of load handling systems capable of carrying loads over the spent fuel pool (regardless of capacity or load carried) is consistent with Phase II of NUREG-0612.

2.2.2 Evaluation of Phase II Compliance

a. Summary of Licensee Statements and Conclusions The Licensee selected Alternative 4 from those identified in NUREG-0612, Section 5.1.2 for the evaluation of identified cranes. However, no analysis, as stipulated in Alternative 4, has been performed. The Licensee believes that the likelihood-of a load drop from the fuel handling crane (L3) and the Unit 1 auxiliary fuel handling crane that would result in an unacceptable radioactivity release is extremely small for the following reasons:

1. Plant technical specifications prohibit the handling of loads in-excess of 2000 lb over spent fuel, and plant procedures have been-revised to prohibit the storage of irradiated fuel in the vicinity of the fuel pool gates and around the periphery of the fuel pools.

j 2. A safety factor of 5 to ultimate was used for the design of the 100/10-ton fuel handling crane.

3. The crane's box girders, as well as the crane girders and their support structures, were designed to resist both design basis and maximum' credible earthquake forces. The crane box girders were also designed to resist vertical earthquake-induced loads with the lifted f load of 100 tons. - The crane girders and support structures were also designed to resist -tornado wind loadings.

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4. The crane is inspected in accordance with ANSI B30.2-1976 prior to r

its use, and the slings used in load handling operations with this j  : crane comply.with ANSI B30.9-1971 requirements.

5. The reinforced concrete slab on the bottom of the spent fuel pools in -

both units has been analyzed for a pool gate load drop in accordance with an impact load analysis methodology contained in ASCE's Report

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on Engineering Practice No. 58, " Structural' Analysis ' and Design of

. Nuclear Plant Facilities,* 1980. This analysis revealed that, although a fuel pool gate would slightly penetrate the bottom slab in the fuel pool, the slab would neither spell nor fail in' shear.- It is-I believed that this' analysis is conservative and therefore an acceptable method of excluding a- fuel pool gate ' drop from further consideration.

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TER-C5506-442/443 The Licensee also believes that the crane's hoist load block can be excluded from further consideration for the following reasons:

1. The main hoist is only used to lif t a spent fuel cask or a fuel handling machine (if necessary) and other miscellaneous loads that are less than 2000 lb. Plant procedures identify load paths that do not cross over spent fuel in the pool.

2. Before use, the crane is inspected per ANSI B30.2, and all slings used with this crane meet ANSI B30.9 requirements.

3. The only crane failure that could cause the main hoist block to fail and thus - fall on spent fuel in the pool would be a "two-blocking" event, which cannot occur with the hoist not in use.

4. The main hoist load block is considered an integral part of the crane.

The Unit 2 new fuel handling crane (Equipment No. 2L35) is used to handle new fuel, new control components, and other miscellaneous loads, such as steel hatch covers at elevation 404 ft in the auxiliary building, that crane L3 cannot reach. The only loads handled over fuel are either new fuel, new control components, or loads less than 2000 lb. The Licensee believes that this erane should be excluded fross further consideration under NUREG-0612 for the following reasons:

1. The crane was designed and constructed per CMAA-70.

2. The crane is inspected and operated per ANSI B30.2-1976, and all slings that are used with this crane are in compliance with ANSI B3 0.9-1976.

3. ' The load of a new fuel assembly and its handling . tool for Unit 2 is approximately 1450 lb, which is 18% of the crane's 4-ton rated capacity and less than the 2000-lb limit of loads carried over spent.

fuel in the pool; A new control component weighs approximately 96 lbr a hatch cover weighs approximately 2300 lb but is not handled

.over the fuel pool.

b. Evaluation and Conclusion  !

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The Licensee has' not provided an analysis per NUREG-0612, Appendix A

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requirements, as stipula'ted in selected Alternative 4 of Section 5.1.2, to

- thow that a postulated load drop into -the spent fuel pool would not cause l

_ uncontrolled release of radioactivity in excess of the dose limits of i 1

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1 TER-C5506-442/443 Criterion I. Instead, the Licensee has relied on technical specification restrictions and administrative controls on the movement of loads over the spent fuel to conclude that the likelihood of load drop is extremely small.

Reliance on safe load paths and other such adminstrative controls in lieu of

, mechanical stops or electrical interlocks does not necessarily ensure that the heavy loads will not be handled over the stored spent fuel due to operator inattention or error. The Licensee's discussion of crane features, other than seismic design, which make the likelihood of a load drop from the spent fuel area cranes extremely small addresses only features that conform to industrial standards and are consistent with the general requirements of NUREG-0612, Section 5.1.1. They are not extra features to improve the reliability of cranes' handling critical loads as discussed in Section 1.2 of this report.

Technical specification restrictions on loads (not to exceed 2000 lb) to be carried over spent fuel stored in pools land much more credibility to the

Licensea's conclusions, and the Licensee has tried to take credit for these r

l restrictions to . demonstrate that a fuel handling accident would.not result in potential offsite exposures in excess of dose limits specified in evaluation Criterion I of Section 5.1 of NUREG-0612. The Licensee, however, has not tddressed the concern with regard to the criticality aspects of Criterion II-following a postulated heavy load drop accident.

f-In conclusion,- it appears that the Licensee has attempted to show conformance.to the Guidelines of NUREG-0612, Section 5.1.2 through a' 2 combination of load drop analysis (for damage to the fuel pool subsequent to a j gate drop) and a technical specification (for damage to spent fuel) . 'Although the use of'a technical specification may be found to provide protection equivalent to mechanical stops or electrical interlocks, as discussed in NUREG-0612, Section 5.1.2(2) ,: such a determination cannot be made on. the basis cf the.information provided.. In order to be found equivalent to electrical interlocks in reducing the probability of a load drop, the associated technical specification should preclude carrying heavy loads a reasonable distance away from the stored -fuel so that the load-cannot be carried over-spent ^ fuel as a result of operator inattention or control system failure and

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so that the fuel will' not be impacted following a handling system failure,

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including an unfavorable handling device failure when the crane is operating in the area permitted by the technical specification. Further, the technical specification should be easily enforceable through markings or other  !

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. provisions that allow an immediate determination of whether a technical cpecification limit has been exceeded. The Licensee should provide additional i

information concerning technical specification limits and enforcement  !

i provisions to allow evaluation of this alternative. ,

2.3 OVERHEAD HANDLING SYSTEMS IN REACTOR VESSEL AREA f

2.3.1 Identification of overhead Handling Systems i'

a.

Sunniary of Licensee Statements and Conclusions  !

1 The Licensee identified the following load handling systems as being .

subject to Phase II criteria of NUREG-0612:

o polar cranes (L2 and 2L2) o control rod drive (CRD) and general maintenance crane (L21) .

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The polar cranes at Arkansas Nuclear One Units 1 and 2 are PER Earnisch-fager 150/25-ton capacity, doublM antry circular cranes. The equipment .

numbers for these cranes are L2 and 2L2 for Units 1 and 2, respectively.

The Unit 1 CRD and general maintenance crane (Equipment No. L21) was '

manufactured by Heco-Pacific Manufacturing Company and is a'2-ton capacity, '

top-riding, single-girder crane consisting of a bridge, monorail, trolley, and - k hoist. The Licensee believes that the CRD and general maintenance crane (L21) l in the Unit i reactor building can 'be excluded from further consideration for I i

tho following reasons: [

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1. Although the crane has a rated capacity of 2 tons, its maximum lif ted i

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load is a CRD mechanism whose total assembly weight is 935 lb. - This ~

. crane also is used to move reactor vessel studs .which weight 640 lb, j coactor vessel head insulation pieces with an average weight of ,

. approximately 400 lb, portions of reactor vessel head cooling duct I w ek whose maximum calculated weight is approximately 800 lb, and j several other small miscellaneous loads. None of these loads is a l heavy load since- their respective weights do not exceed 2000 lb.

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2. This crane is used to assist in several maintenance operations prior to the removal of the reactor vessel head. However, administrative l controls are being developed to ensure that, prior to the removal of l t the reactor vessel head, the crane is locked in a position at the east end of the refueling canal so that it is incapable of carrying  ;

any load over the open reactor vessel. Administrative controls are '

4 being developed to ensure that it is also locked in this position and j seismically restrained during normal plant operations.

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Also, administrative controls are being developed to ensure that the trolley and hoist are removed from the crane gantry and stored elsewhere during normal plant operations since there are no seismic restraints on the trolley.

b. ~ Evaluation and Conclusion i

The Licensee's identification of load handling systems capable of carrying loads (regardless of capacity or load) over the reactor vessel is d

consistent with NUREG-0612.

a The Licensee's justifications for the exclusion of the CRD and general maintenance crane (L21) meet the intent of NUREG-0612 contingent upon implementation of intended administrative controls.

2.3.2 Evaluation of Phase II Compliance s

a. Summary of Licensee Statements and Conclusions No analysis was performed in accordance with the guidelines of NUREG-0612, Appendix A to demonstrate compliance with Criteria I through III.

However,-a head drop analysis from 3 ft.6 in above the vessel flange, performed by Babcock and Wilcox in November 1972, revealed that the support Ekirt assembly will be generally overstressed in compresgion and will yield moderately or buckle. A total displacement was estimated to be 7/16 in. This analysis is conservative .since it was assumed that the head was a. rigid

' falling weight and no attempt was made to evaluate its own elasticity which would have reduced the stresses even further.

The Unit 1 and Unit 2 polar cranes (Equipment.Nos. L2:and~2L2, respec-

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tively) are rated at 150 tons for their main hooks and 25 tons for their

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cuxiliary hooks. Both cranes were designed.to support the maximum loads

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l caused by a 600-ton construction lift as well as normal operating loads plus l earthquake forces resulting from a design basis earthquake. l t

In susmary, it is believed that the likelihood of a load drop from the l

polar cranes that would result in an unacceptable radioactive release is j extremely small for the following reasons:

1. A safety factor of 5 to ultimate was used for the design of the 150/25-ton polar cranes.

2. The box girders and the crane girder and its brackets were designed for a 600-ton (605 tons on Unit 2) load, and the maximum normal load is only 25% of this on Unit 1 (24% on Unit 2) .

3. The polar cranes are inspected in accordance with ANSI B30.2-1976 only during outages. All polar crane inspections (except daily inspections) are performed under plant surveillance test programs when they are due. Surveillance that is due during plant operation will be performed at the beginning of each outage prior to the use of the crane. This includes periodic, quarterly, semiannual, and annual inspections. Plant procedures are being revised to ensure that the daily (frequent) polar crane inspections are performed prior to the use of these cranes. Exceptions to this are short-lived outages that do not require use of polar cranes. Slings used in lifting heavy loads will be in compliance with ANSI B.9-1971, and crane operators are trained and qualified in accordance with ANSI 830.2-1976.

4. The cranes were designed to resist earthquake forces generated by the design basis earthquake and the maximum credible earthquake.

The in-service inspection (ISI) tools are attached to the polar crane cuxiliary hook and are considered as individual loads. The ISI tool is used

^ for the reactor- vessel inspection and the upper nozzle inspection on Unit I with fuel in the core and the upper plenum assembly removed. The ISI tool has design features that preclude the inadvertent contact of the tool's mast and remote arm with the irradiated fuel in the vessel core. This prevents an cccidental criticality of- the fuel which could be caused by the fuel being -

crushed. :The ISI tool is not used on Unit 2 with fuel in the reactor. Based _

on the above considerations, the Licensee believes that the main hoist load block on polar cranes L2 and 2L2 should be excluded -from further consideration cs a heavy load.

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b. Evaluation and Conclusion The Licensee stated that the likelihood of a drop from polar cranes is extremely small because of the safety factors involved in the design of the cranes as well as improved inspection, maintenance, and operation procedures.

The provisions are consistent with the general requirements of NUREG-0612, Section 5.1.1. They are not extra features to improve the reliability of the cranes handling critical loads as discussed in Section 1.2 of this report.

Pcc the load drop scenario analyzed by Babcock and Wilcox, the structural integrity of the reactor vessel is apparently maintained, although local overstressing and buckling of support skirt occurs following a reactoe vessel '

head drop from a height of 3.5 ft. The Licensee, however, did not provide the rationale for selection of the drop height employed in the reactor vessel impact analysis. The staff's intent, as expressed in NUREG-0612, Appendix A, was to ensure that an analysis relied upon to demonstrate the acceptability of the consequences of a load drop properly considered the maximum height f. on which the load might be dropped. The Licensee should provide justification for the selection of a 3-ft 6-in drop height including an indication and justification of the margin between this height and the maximum height that the reactor vessel head will be carried over the vessel. The margin selected should be reasonable and should accommodate both operator response and the ability of the operator or supervisor to ensure that the margin is not exceeded.

In addition, the Licensee's load drop analysis has not directly addressed whether or not Criteria I through- III of NUREG-0612, Section 5.1 will be satisfied following the postulated load drop. Although it may,be reasonable to assume that an accident that results in reactor vessel displacement of 7/16 inch will not result in fuel damage sufficient to cause an excessive radioactivity release og increase in reactivity, .such conclusions should be drawn by the Licensee on the basis of its analysis and cannot be arrived at-independently on the . basis of the information provided.

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J TER-C5506-442/443 2.4 OVERHEAD HANDLING SYSTEMS IN AREAS CONTAINING SAFE SHUTDOWN EQUIPMENT 2.4.1 Summary of Licensee Statements and Conclusions The Licensee has provided detailed information in matrix form which identifies all handling systems, locations, and loads that may affect equipment or components required for safe shutdown; this information is provided in Appendix A. Discussions have been provided to eliminate load / impact area combinations (Hazard Elimination Category) based on separation and redundancy of safety-related equipment at i other site-specific considerations. These considerations are summarize.J bet.

Intake Structure Gantry Crane (L7)

The heaviest safety-related loads lif ted by the intake structure gantry crane (L7) are the service water pumps, and it is not physically possible to drop one pump on top of another due to the separation of the pumps. This ceparation consists of thick, reinforced concrete walls between each pump bay in both Units 1 and 2. The Unit 1 motors are approximately 27 ft apart with 18-in-thick reinforced concrete walls separating' them. The Unit 2. motors are 4 epproximately 8 f t apart with IS-in-thick reinforced concrete walls separating them. In addition, the size of the motors allows them only to be lifted i through the' roof hatch directly above them.

At present, there are no administrative controls to prevent the handling of other loads at the intake structure, such as the circulating water pumps, ,

ever safety-related equipment in this area. Intake structure crane procedures cre being revised to incorporate these restrictions.

Fuel Handling Crane (L3)

Other than the spent fuel shipping cask, whose load drop has already been analyzed, the fuel handling machine is the heaviest load lif ted by the fuel handling crane (L3). This load weighs less than the 100-ton shipping cask,

, ind the slab at elevation 404 ft 0 in has been shown to survive a cask drop; therefore, the crane can be eliminated from further consideration because a l load drop would not prevent safe reactor shutdown (7].

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P TER-C5506-442/443 It should be notad that there are no safe shutdown components beneath the load paths of the remainder of the heavy loads handled by the fuel handling crane. The crane load block has been previously excluded from further consideration based on a discussion in Section 2.2.2.a.

Polar Cranes (L2 and 2L2) ._

The Unit 1 and Unit 2 polar cranes (Equipment Nos. L2. and 2L2) can be eliminated froin further consideration due to separation and redundancy of safety-related Sipment.

In Unit 1, the heaviest load that would be lifted during any plant condition other than celd shutdown would be the reactor vessel missile shields. The missile shields are normally lifted when the plant is in cold shutdown. Lif ting these missile shields during hot shutdown would be extremely unusual. The worst-case load drop of a missile shield would affect a portion of decay heat removal system piping at the west end of the refueling canal at elevation 354 ft 0 in. This postulated case has been excluded from further consideration because the piping in question is routed against the

- outside face of the secondary shield wall and the load would have to slide down the shield wall in order to destroy this piping. Due to the physical orientation of the missile shield, it has been determined that this is an incredible scenario.

When the plant is in a cold shutdown condition or in a refueling shutdown condition, the remainder of the heavy loads listied in Appendix A may be lifted. There are only four loads whose load drop could affect the ability to maintain the plant in a cold shutdown condition, i.e., maintain decay heat reseval capability. These are the reactor. vessel head, the reactor coolant pump motor, pump, and structural support beam. The load paths of the reactor

. coolant pumps located in the north cavity pass over one decay heat removal line located near-the reactor building sump. A load drop onto this line would not result in the loss of decay heat removal capability because one decay heat loop would still be^available. The load paths of.the reactor coolant pumps located in the south cavity pass over the "A" core flood line on the west side -

of the secondary. shield wall, but a load drop on this line would not result in pw OUuu Franken Reneerch Center A Onuman of The Pasmi enmunne

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I I TER-C5506-442/443 the loss of decay heat removal capability. Finally, the load path of the reactor vessel head passes over the core flood line and the decay heat removal line for the "A" loop. Even if these lines were destroyed by a head drop, the rcdundant decay heat resoval loop could still maintain core coverage. ,

In Unit 2, the heavy loads that might be lifted by the polar crane in a plant condition other than cold shutdown would be the vessel head stud stand or the refueling canal seal plate lifting rig. They are normally lifted when the plant is in cold shutdown. Although these loads pass over safety injection piping, the possibility cf these loads penetrating several feet of roinforced concrete is extremely remte. For this. reason, the carrying of ,

these loads under a plant condition other than cold shutdown can be eliminated from further consideration.

When the plant is in a cold shutdown or a refueling shutdown condition, the remainder of the heavy loads listed in Appendix A may be lif ted. There are several loads that may be lif ted whose load drop could affect the ability to maintain the plant in a cold shutdown condition, i.e., naintain decay heat {

eceval (shutdown cooling) capability. These are the reactor vessel head f lift, the reactor coolant pumps and their structural steel support beams, the hcad maintenance structure, the jib crane (2L45), the refueling machine and i

other refueling system components, and the refueling cavity seal plate. Of  ;

these loads, the consequences of a reactor coolant pump motor drop or a , l l

ccactor vessel head drop would envelope the other postulated load drops. ,

A reactor vessel head drop over the refueling cavity could result in the l 1 css of shutdown cooling and safety injection piping from the reactor vessel f

through the "B" bot leg. However, this would only occur if the head penetrated j c 4-f t-0-in-thick concrete slab. This would not cause loss of decay heat ecmoval capability from the reactoe vessel because this line could be isolated l I

cnd the other cooling loop could be used to maintain shutdown cooling.  ;

A drop of the "A" reactor celant pump motor over the reactor building sump area could result in the loss of shutdown cooling and safety injection piping . into the "C" cold leg as well as one containment spray line. It is  !

Uh ranklin Research Center A Osamen of The Frereen basemme

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TER-C5506-442/443 still possible to maintain adequate core decay heat removal after the loss of this piping by placing the shutdown cooling system in the recirculation mode l using the HPSI pumps and by isolating the affected safety injection line.

Main Steam Isolation valve (MSIV) Bridge Crane (2L10) j The load lifted by the MSIV bridge crane (2L10) and its respective load / target combinations can be eliminated from further consideration because a load drop would neither prevent safe reactor shutdown nor prohibit continued decay heat removal. The reactor and the plant must be in a cold shutdown condition before a MSIV is removed so a load drop would not prevent safe reactor shutdown. The heaviest component of a MSIV lif t is a cylinder 4

stiffener section that weighs approximately 10,000 lb. It is postulated that '

this load, if dropped, would penetrate the north penetration rooms which contain piping and electrical conduit servicing the service water system, the emergency feedwater system, the fire water system to containment, the main feedwater system, the MSIVs, penetration room ventilation, and a containment HVAC radiation monitor. Since the loss of these systems or portions of them ,

when the plant is in a cold shutdown condition will not prohibit continued decay heat removal, this crane can be eliminated from further consideration. ,

2.4.2 Evaluation and Conclusion Information provided by the Licensee is insufficient to allow an' ,

. independent determination that all equipment associated with plant shutdown or decay heat removal in the vicinity of paths followed by heavy loads have 'been o

evaluated. The Licensare should provide additional information showing the

i. location of load paths and equipment, including cables and motor control centers, associated with plant shutdown and cooldown. Additional information is also needed to justify various hazard elimination categories in Appendix A.

i sat as

' UIJ ud FrenWin Research Center A Chagen af The Fangen buense

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TER-C5506-442/443

3. CONCLUSIONS This annemary is provided to consolidate the results of crane-specific evaluations presented in Section 2. It is not meant as a substitute for the cpecific conclusions reached in the various subsections of Section 2. It is provided to allow the reader to focus on the key topics which should be cddressed in seeking to resolve issues where the degree of load handling reliability provided by cranes at Arkansas Nuclear One Units 1 and 2 was not found to meet the objectives of NUREG-0612. This section addresses issues for which the information provided is felt to be inadequate to support a definitive conclusion and issues wherein the information provided has been evaluated as proposing an approach inconsistent with the guidance of NUREG-0612.

l 3.1 INFORMATION ISSUES The following information provided by the Licensee has been assessed to I

be insufficient to support an independent conclusion that load handling roliability is consistent with the evaluation criteria of Section 2.1 in the -

following areas: [

Reactor Vessel Area Ioad Handling Systems (Section 2.3.2) j o The Licensee should provide justification for the selection of the l 3-ft 6-in drop height, including justification of the margin between l this height and the maximum height that the reactor vessel head will {

be' carried over the reactor vessel. The analysis should be performed ,

in accordance with NUREG-0612, Appendix A and based on maximum fall  !

f with due allowances for operatoc errors.

o The Licensee shoulJ indicate that provisions that are in effect to 4- prevent exceeding the procedural maximum lift height.  ;

o The analysis performed should indicate satisfaction of evaluation Criteria I through III of NUREG-0612, Section 5.1.

I Safe Shutdown Equipment Area toad Handling Systems (Section 2.4.1) l o The Licensee should provide additional information showing the l location of load paths and equipment, including cables and MCCs, t associated with plant shutdown and cooldown. l i

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-L6-1 i Lu Frankhn Renesch Center 4c w n. r so j

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TER-C5506-442/443 l o The Licensee should provide additional information to justify various

)

hazard elimination categories in Appendix A of this report.

3.2 APPROACH ISSUES This review has revealed one issue wherein the approach or position taken by the Licensee, based on information provided thus far, is inconsistent with i the staff's objective as expressed in the evaluation criteria of Section 2.1:

Spent Fuel Pool Area Ioad Handling Devices (Section 2.2.2) o The Licensee appears to rely on the use of technical specifications and administrative controls to eliminate from further consideration

certain heavy loads handled in the vicinity of the spent fuel pool.

In general, such procedural controls are not equivalent, in accordance with NUREG-0612 guidelines, to physical restraint or enhanced load handling system reliability in reducing the likelihood of a load drop 1 over spent fuel. It is recognized, however, that in certain unique circumstances (specifically where the administrative controls provide large separations between the control limits and the impact area of interest that are readily monitocable and strictly enforced),

administrative controls can be found, on the basis of engineering judgment, to provide a high degree of certainty that loads will never be carried over the target. The Licensee has not demonstrated that

. these restrictions exist or that their exception is appropriate.

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I Ou' u Frankhn Reneerch Center AOsamenof Die Frusueninsensee ,

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TER-C5506-442/443

4. REFERENCES

1. V. Stello (NRC)

Letter to All Licensees

Subject:

Request for Additional Information on Control of Heavy Loads Near Spent Fuel May 17, 1978

2. NRC NUREG-0612, " Control of Heavy Icada at Nuclear Power Plants

• July 1980

3. D. G. Eisenhut (NBC)

Letter to All Operating Reactors

Subject:

Control of Heavy Inads December 22, 1980

4. FRC Technical Evaluation Report, " Control of Heavy Ioads at Arkansas Nuclear One Units 1 and 2" TER-C5257-59/531, December 21, 1981

5. John R. Marshal (APEL)

Letter to D. G. Eisenhut (NBC)

Subject:

Control of Heavy Loads (Phase II)

December 22, 1982

6. NRC NUREG-0544, " Single-Failure-Proof Cranes at Nuclear Power Plants" May 1979

7. D. H. Williams (APEL)

Letter to R. W. Reid (NBC) i Subject Cask Handling (Cask Drop Analysis)

July 19, 1978 Ag' O'u u d Frankhn Reneerch Corner A Duman af The Pieuman m

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APPENDIX A ICAD/ IMPACT AREA MATRIX, ARKANSAS NUCLEAR ONE UNITS 1 AND 2 '

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b. . . Franklin Research Center A Division of The FranFJin Institute The Bensmen Frenidn Parkwey, Pheia . Pe 19:03(21Si448 1000

l Appendia A. Load / Impact Area Matrix, Artansas Nuclear One Units 1 and 2 Impact safety-Related Hazard Stim.

Crane tocation Elevation toad Impact Area F4u isument Category

• Fu;l Mandling Reactor Ausillary 404'-0 Spent fuel Control hom Ikaof Relay Panels E Crane (t.3 3 eldg. (ruel Mandl- Cast (25-ing Areal Ton) 404'-0 Match over Match Frame None C Crano say 306'-4 ruel Trane, ruel Tilt Pit rioor None C Tube Gate valve 362'-0 New ruel ruel Tilt Pit Floor None C Elevator 404'-0 Fuel Hand. Control Room ftzot Relay Panels E Machine .

362'-0 Uponder Fuel Tilt Pit ricor None C 354'-0 New ruel R. R. eey rioor None C Ship. Cont.

354'-0 New Control R. R. Bey Floor None C Comp.

354'-0 New ruel R. R. Say Floor Hone C Elseents 354'-0 New Control R. R. any rioor None C equipeent 162'-0 Fuel Transfer ruel Tilt Pit rioor None C Carriage 362'-0 ruel Pool Spent ruel Pool ruel Pool E Divider Gates rioor 404'-0 Crane toad ruel Mand. Floorp Spent ruel in D Block ruel pool ricors ruel Poolt Relay R. R. Say rioor Panels in Control Room I

• %ianetton of Hasard Elimination Categoriest A = Crane traret for this area / load cosmination prohibited by electrical interlocks or machanical stops.

E e Systee redundancy and separation precludes toes of capability of 'systee to perform its safety-related function following this load drop in this area.

C e Site-specific consideratione elietnate the need to consider load / equipment commination.

0

• Liteithood of handling system failure for this load is entremely small (i.e., Section 5.1.6, WREG-0612 gatisfiedt .

E e Analynig deenstrates that erane f ailure and load drop will not damage safety-related equipment.

• A-1

[dNU Frarddin Research Center 4 Dammen of The Fromen insense s e a. .h

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0 Appendix % (Con t. )

tapact Safety-Ralated Hazard Elsa.

Crsne tocat ion Elevation Load tecoct Area toutpoent Category

• New Fuel Reactor Auxiliary 404'-0 New Fuel Control lboe Ibof g Relay Panela in E Randling Crane Oldg. Unit 2 or Asaeebly Spent Fuel in Pool Control feoes (2L11) 362*-0 Spent fuel in Pool 404'-0 New Cont. Control fboe Ibof f Relay Panels in 8 or Component Spent Fuel in Pool Control Roomt 362'-0 Spent fuel in Pool Int:te Intate Structure 364'-0 Unit 1 Service Water Pump Pump Room 8 l Structure Unit 1 or Service Water !boe Floor or intate i centry Crane 378'-0 Pump pector Structure Ibot 1 (L7) 322'-6 Unit 1 Intate Structure Pump fbom 8 or Service Water Base Met 378'-0 Pump 364'-0 petor Driven Pump Room Flwr or Pump Iboe 8 or Fire Pump Intate Structure Floor 378'-0 and letor Ibo t 366'-0 Diesel Driven Pump Room Floor or Pump phoe 8 or Fire Pump Intake Structure Floor 178'-0 teof r 166'-0 Jockey Pump Pump Room Floor or Fire Pump 8 or Fire Pump Intate Structure 378'-0 Ibo f L

364'-0 Unit 2 Service Water Pump Pump Iboe 8 Intate intate Structure Structure Unit 2 or Service Water Intate Room Floor or Gantry Crane 378'-0 Pump bestor Intate Structure Roof (L7)

Unit 2 Mone 8 322*-6 Intate Structure Service laater Saee %t Pump 370'-0 Room Match intate Structure Pump leos 8 IItate intate Structure Structure Unite 1 and 2 Pluge Roof cantry Crone 3 (L7) 1 a

1. b ~A-2 il FranMn Research Center A b ef The Framen henne .

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Appendia A (Cont.)

Impact Safety-Related Hazard Elim.

Elevation toad tapact Area equipment category

• Crane toestion Unit i Polar Reactor sidg. 426'-6 Missile Top of secondary Reactor Vessel 5, C Cetne (L28 Unit i Shielda Shield ten 11s Headt RCPs "A* 4 ,

"C* Pressuriser Steam Generatorst >

Letdown Pipings Decay Heat Removal Piping  ;

176'-6 Top Head Refueling Cavity Reactor Vessel 8 Insul. with Floor Head w/ Storage pactsb 362'-0 Transfer Tube Refueling Cavity None C Flangeb Floor 376'-6 Reactor Vessel Refueling Cavity Reactor Vessels B 401'-6 Heedb F1Jorg Head Stands Decay Heat or Equipment Hetch Removal Piping 357'-0 Area 362'-0 Upper Plenum Reactor Vessel Reactor Vessels e i Assemelyb Refueling Cavity Fuel in Vassel l 6

Floor 376'-6 Stud Storage Refueling Cavity None C RactU Floor 376'-6 Indexing Refueling Cavity or Reactor Vessel 8 Fintureb Reactor vessel Flange 376' 6 Head and Reactor Vessel Head Reactor Vessel 8 Internal Refueling Cavity or Headt RCP "A's Handling Head Stand 'B' Steam Gener-Fisture Lift ators Core Flood Rig w/ Turn. Piping buctiesb t

a. toada eith this desigaation can only be moved around the containment at elevation 424'-6* (426'-6* on Unit 2) . - These loads will be moved as close to the top of the secondary shield walls as possible with adjuetments '.c the load's clevation as required to avoid such obetructions as handrails. Unit 2 main steme lines, etc.

D. T%is designation on the table indicates that these loads sove over the reactor cavity.

t ranklin Research Center A Chuman af 7he Fressen insumme

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Appendix A (Con t. )

Impact Sa f ety-Related Hasard Elim.

Cran 7 tocation Elevation toad faoect Area Etruipment Category

• rMit 1 Polar Reactor Bldg. 376'-6 Refueling Refueling Cavity or RCP *A* or "S*r 5 Crtne (L2) Unit 1 Cavity Reactor Vessel Reactor Vessels (Cont. ) Seal Plateb one Steam Gener-ator: Decay Heat Receval & Core riood Piping 357'-0 Unassembled Equipment Hatch Area ' A' Cold Leg,

• A* 8 401'-4 ISI (ARIS) Operating Floor South RCP, 'B' Hot Legs or Tool cavity *3* Steam Gener-424'-4 ators Core riood Piping 176'-6 Assembled ISI Top of *D* Ring, *3* Not Leg and B 424*-4 (AAIS) 10o1D Reactor vessel Steam Generator Reactor Vessel 336'-4 RCP m tor of 81dg. Basement or RCP HPI Nossles 5 or Pump a Equipment Batch in Cold Leg 357*-0 Reactor aldg.

Sormy needer:

Decay Rest Removal Piping and Core riood Piping 334'-4 Structural Same as EP mtor Same se RCP Motor 8 or Beame Above or Pump or Pump 357'-0 KP *A*, *D**

334'-4 Structural Same as RCP Motor Same as RCP Motor a or Beame Above or Pump or Pump 357'-0 RCP 'S*, *C*a 336'-6 ruel Transfer 31dg. Basement HP1 Mossles *A* 8 357'-0 Carriageb Equipment Matcht 6 *3* Reactor 362'-0 operating rioor Vesselp *a" Steam 176'-6 Top of "D* Generator 401'-6 Ring (South),

or Refueling Cavity 424'-6 rioor t

I d% A 1

{iNU Franklin Research Center a Domme w ne r,meen = l

.. . . . _ . . . . . . - j

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Appendix A (Cont.) i i

l Impact $4fety-Related Hatard Elim. f cr'ne tocation Elevation toad Insect Area Equipment category' l finit 1 Polar Reactor 81dg. 336'-4 ' 09enderb 31dg. Basement; HP! Nosslee *A" 8 Cecn3 (f.2 3 Unit 1 357'-0 ' Equipment Maten 6 *2's Reactor  ;

(Cont.V 362'-0 Operating Floors vessels "B' State  ;

376'-6 Top o f *D* Generator j 401'-4 eing (Southlt or Refueling Cavity 424'-4 riaor 376'-O' Ra?celing 51dg. Basements HPt Nossise *A* 8 357'-0 Maciine equ3pcent Ratcht 6 *B's RCPs *A* i 424'-6 ,.

Operattng Floort 6 *s' Steam J

- Top of *D* Generator Ring (South):

Refueling Cavity b Floor 376'-0 Auxiliary 51dg. Basements .HPI Nossles /A* 8 357'-0 Retailing. Equipment Natcht 6 *B*3 RCPM *A* i 424'-4 Me@ine Operating Floorg - 6 *5* State

. Top of *D* Gene 4ator  !

Ring (south) 1 liefueling Cavity  !

Floor i

376'-0 Refuetius sidg. assement HPI Norales 'A' S 357'-0 Canal equipment Matuhr 6 *B*r RCPe "A* r 424'-4 .tAdderb Operating Floor; & *B* Steam

• top of *O* Generator Ring eSouthl Refueling Cavity-- i Floor Any Slov. Crane-toed Any Area Any Safety-Related O

s. ' alocar (Main. Equipment under a Hola tib toad Path Unit 2 Polar Reactor 81dg. 365'-0 Head poteeling, Cavity Reactor .Versel 8 [

Crane (2t.231 unit 2 426'-4 te iinteet ence riover yop of Head Presourizers 3true tureD Secondary Shield RCF *A*g RCP *C's I

. idelle sefety injestion of Shutdowie Cool-(. y ,

< -j , ing Piping

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Appendix A (Con t. )

Impact safety-Related Hazard Elim.

Crine tocatton E1evatton toad Imonet Aese Eau 1oment category

• Unit 2 polar Reactor sidg. 365'-0 Head Refueling Cavity Reactor Vessel 8 Crrne (2L2) Unit 2 357'-0 CEDM Refueling Cavity All RCPs: Both 8 (Cont.) 376'-0 Ductworg b Ploor Top of Steam Generators:

426'-6 Secondary Shield Safety Injection walls Equipment tanta "A* 4 "D*

Ratch 376'-0 CEst Cooling Refueling Cavity All the Above for 8

! 426'-6 Shroude Ploor Top of CEDM Ouctwork Plue Secondary Shield 'B' Hot Leg ,

4 wall 336'-6 RCP Motor Saeement Top of Respective RCP B 426'-4 and Pumpa Secondary Shield Pumps Pressuriser 357'-0 Walls Equipment surge Lines Safety Natch Area Injection and Shutdown Cooling Piping Cold Lege Steam Generatoee 334'-4 Structural sasements hp of Respective EP S 426'-6 sease Above Secondary Shield Pumps Pressuriser 357'-0 RCP 'A' te walls Equipment Surge Lines Safety

• 3.a Hatch Area injection and shutdown Cooling Piping Cold Lege Steam Generators 336'-4 Structural Basements hp of Respective RCP 8 426'-4 Seeme Above Secondary Shield Pumps Pressuriser 357'-0 KF *C* & walls aquipment Surge Line Safety

• D'a Ratch Area Injection and shutdown Cooling Piping Cold Lege Steam Generatore 336'-6 Inservice sesement Equipment RCP *C*: Reactor 8-357'-0 Inspection Hatch Areas Refuel- Vessel Safety

. -376'-0 Molab ing Cavity Ploor Injection and l -405' 6 Operating Ploor Shutdown Cooling

! 4268 4 2p of Secondary Piping into 'C' l Shield well Cold Leg e

4

- .3,s.e, e -

Appendi#. A (Cont. ) '

Impact safety-Related Hazard Elim.

Crane tocation elevation. toad Impact Area Equipment Categorva Unit 2 Polar Reactor aldg. 354'-0 ruel Trenafer Equipment Match Areap *A* e *B" 8 Crane (2L2) Unit 2 362'-0 Carriage D Refueling Cavity Cold Legs *A*

(Cont.) 405'-6 rioors operating Rot Legs Safety 426'-6 rioort Top of Injection and Secondary Shield Shutdomi Cooling wall Piping 354'-0 Uponder b Equipment Ratch Areap *A* & *B' S 162'-0 Refueling Cavity Cold Leg; *A*

405'-6 rioor Opera ting not Leg: Sa f ety 426'-6 ricors Top of Injection and secondary Shield Shutdown Cooling well Piping 354'-0 Rafaelint Equipment Hatch Area *A* & *B" B 362'-0 Machine Refueling Cavity Cold Legs *A*

405'-6 rioors operating Hot Legg Safety 426'-6 '

Floors Top of Injection and Secondary Shield Shutdown Cooling Wall <

Piping 376'-0 Stud Tension Refueling Cavity teone C 405'-6 and Hydraulic ricore operating unitD rioor 357'-0 Reactor Equipment Match Areas Reec or vessel e i 376'-0 vessel Readb Refueling Cavity Safety Injection f 365'-0 rioors operating and Shutdown Cool- I rioer ing PipLng i 405'76 357'*0 Reactor Equipment Match Areat Reactor Vessel S 376'-0_ Vessel Head Refueling Cavity Safety Injectisi 365'-0 Lift Rig h rioors operating and Shutdown Coci-405'-6 rioor ing Pipino 405'-6 Reactor operating rioor teone C ves:e1 Itered ,

Studs 376'-0 Upper Guids. Refueling Cavity Reactor Vessel B Structureb/ rioor Internals ruel in Core ,

376'-0 '#pper Guitie Same as Upper Guide Same as Upper 8 Structur e Structure Guide Structure Lift Rigb 176'-0 Cora Barreib Same as Upper Guide teone C Structure

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1. % A-7 N UkFranklin Research Center  ;

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Appendix A (Cont.)

Impact Safety-Related Hasard Elim.

Crane tocatton Elevatton toad Imonet Ares Eau 1poent cateaory*

Unit 2 Polar Reactor aldg. 376'-0 Core secrel same as Upper Guide tone C Crane (2L2) Unit 2 Lift Righ Structure (Con t. )

336'-6 Jib Crane Basement Floor Reactor Vesselp C 357'-0 (2L45) Read Equipment All RCPer Both 405'-6 Natch Areat Operating Steam Generatorer 426'-6 Floors Top of Shutdown Cooling Secondary Shield Piping wall 376'-0 Stud Stand Refueling Cavity None C 405'-6 w/Studab Floor Operating Floor 357'-0 CEDM Equipment Match Areas Reactor Vessel 8 376'-0 Eutension Refueling Cavity Head Pressuriser 405'-6 Shaft Floorg Operating Surge Line to "A" 426'-4 Uncoupling b Floort Top of Rot Legg Shutdown Secondary Shield Cooling Pipings wall *A* Cold Leg 376'-0 Alignment Refueling Cavity None C Pined Floor Vessel Flange 376'-0 Refueling Refueling Cavity Reactor Vessel B 405'-4 Cavity Floort Operating Head Shutdown Seal Plateb Ploor Cooling Pipe to

• 5* Not Leg and

• C* Cold Leg 365'-0 Seal Plata Refueling Cavity Same as Seal Plate 8' 405'-4 Lift Righ Floor Operating 426'-4 Finors Top of -

Secondary Shield well Any 81ev. Main noist Any Area Any Equipment D toad Blocub Below Crane MSiv eridg1 Turbine Auxiliary 404*-0 Main Steam Piping and Electrical Piping and Electrical a Crtne (2L10) 81dg. 386'-0 teolation Penetration Ibom Cablee in North 354'-0 Valve (10,000 Roof - Penetration Phone ibs1

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