ML18152A094

Control of Heavy Loads-Phase Ii,Surry Power Station Units 1 & 2, Draft Technical Evaluation Rept

ML18152A094
Person / Time
Site:	Surry
Issue date:	05/14/1984
From:	Walter R FRANKLIN INSTITUTE
To:	Singh A NRC
Shared Package
ML18141A723	List: ML18152A094
References
CON-NRC-03-81-130, CON-NRC-3-81-130, REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR TAC-52276, TAC-52277, TER-C5506-508-5, TER-C5506-508-509-DF, TER-C5506-508-509-DR, NUDOCS 8406140286
Download: ML18152A094 (30)

Text

9 (DRAFT) 6 TECHNICAL EVALUATIO~REPORT CONTROL OF HEAVY LOADS -PHASE II VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 NRC DOCKET NO. 50-280, 50-281 NRC TAC NO. 52276, 52277 NRC CONTRACT NO. NRC-03-81-130

-._ Prepared by Franklin Research Center 20th and Race Streets Philadelphia, PA 19103 Prepared for Nuclear Regulatory Commission Washington, D.C. 20555 May 14, 1984 FRC PROJECT C5506 FRCAS5IGNMENT19 FRCTASKS 508, 509 Author: R. J. Walter FRC Group Leader: I. H. Sargent Lead NRC Engineer:

A. Singh 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, ratus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights. ~nklin Research Center A Division of The Franklin Institute . 20th and Race Streets. Phila .. Pa. 19103 ((215) 448-1000 e CONTENTS Section Title l 2 3 INTRODUCTION 1.1 Purpose 1.2 Generic Background.

1.3 Plant-Specific Background.

EVALUATION.

2.1 Evaluation Criteria 2.2 Containment.Load Handling Systems. 2.3 Fuel Building Overhead Handling Systems 2.4 Overhead Handling Systems in Areas Containing Safe Shutdown F.quipment

• CONCLUSION.

3.1 Information Issues. 3.2 Approach Issues 4 REFERENCES

• ~nklin Research Center A Olvisian ol The Franldin lnstibile iii TER-CS506-508/509 Page l 1 1 2 3 . 3 . , 4 6 12 24 24 25 26

• TER-C5506-508/509 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. B. Sargent and Mr. R. J. Walter contributed to the technical preparation of this report through a subcontract with WESTEC Services#

Inc * * ~nklin Research Center A Oivision of The Franldin lnslilule

• V

\ e TER-CSS06-508/509

1. INTRODUCTION 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 reactor shutdown and fuel element decay heat removal at Surry Power Station Units land 2. This review constitutes the second phase of a two-phase review instituted 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 Nucl-ear Regulatory Commission (NRt) 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 the.NRC staff on May 17, 1978 (1] to all power reactor licensees, requesting information concerning the control of heavy loads near spent fuel. The results of Task A-36 were reported in NOREG-06.12 (2]. The staff concluded from this evaluation that existing measures to control the handling of heavy loads at operating plants provide protection from certain potential problems but do not adequately cover the major causes of load handling accidents and should be upgraded.

To upgrade measures for the control of heavy loads, the staff developed a series 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 syste~~ at nuclear power plants have been designed and are operated so that their probability of failure is appropriately small for the critical tasks 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 ~nklin Research Cente: A OMsion ol The Franklin institute e TER-CS506-508/509 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 load drop extremely small or (2) conservative evaluations of load handling accidents indicate that the potential consequences of a load drop are acceptably small. 1.3 PLANT-SPECIFIC BACK~OUND On December 22, 1980, the NRC issued a letter [3] to Virginia El-ectric I and Power Company (VEPCO); the Licensee for Surry Power Station, requesting the review of provisions for handling and control of heavy loads, the evaluation of th~se provisions with respect to the guidelines of NUREG-0612, and the provision of certain additional information to be used for an independent determination of conformance to these guidelines.

The results of this independent evaluation with respect to general load handling equipment and procedurea (Phase I) were provided on January 14, 1983 [4]. On March 22, 1982, VEPCO provided an initial Phase II report [5] concerning conformance with 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. I I * * / ~nklin Research Center A Division of The Franillin llmmll! *

• TER-CS506-508/509

2. EVALUATION This sectt::.iam JP)II:esents an evaluation of critical load handling areas at Surry Power Stt:a.ticalm Units 1 and 2. Separate subsections are provided to identify the ciri~ia used in this evaluation and each of the plant areas considered.

!'er each such area, relevant ioad handling systems are identified, L:iicr:enwee-provided information related to the evaluation criteria or proposed aJL.termatives is summarized and evaluated, and a conclusion as to the extent of ci:,:amw,11.iance, including recommended additional action or requirements f<<Jr a&d'ditional information as appropriate, is provided * . 2 .1 EVALUAfl:tiDII CID'J?EIUA

• The objectl:.i"ft!

of this review was to determine if plant arrangements and load handlingJ equuipnent design were such-that either the likelihood of a load handling a~ ttllrat could damage spent fuel or equipment used in reactor shutdown or fmleJL element decay heat removal is extremely small or that the consequences cf sum::tl1 damage, should it occur, will be acceptable.

Guidance contained im NlllllilBIG-ill12, Sections 5.1.2, 5.1.3, and 5.1.5 (for pressurized water reactcirs))

amml in 5.1.4 and 5.1.5 (for boiling water reactors) forms the basis for the camc:U1111Sions reached in this section and is briefly summarized as follows. For a ~ion that the likelihood of damage is extremely small: o The alem:ii.p of: the load handling system (i.e., crane or hoist and underl!DDlmk .lLifting devices) is consistent with, or equivalent to, the NRC sbfif cciteria for single-failure-proof cranes identified in Ntm!)l;HDSSI ff 6] , or o The pi~ ~sical arrangement is such that a crane operated in the vicilmitw elf spent fuel or safety-related equipment is prevented from traveJJ]irn,gi t:o a position from which a load drop can be expected to d~ smdln equipment.

For a ~on that the potential consequences of dam4ge following a lo~d drap vill.]_ he acceptable:

o In t:lhe CillSiie of potential damage to spent fuel, calculations have been provitd!fell 1bD> demonstrate that potential radiological doses at the site boUDlilaur:y v:ii.J!.1 not exceed 251 of the limits specified in lOCFRlOO and *

• e

• TER-C5506-508/509 that the post-accident configuration of the fuel will not result in a Xe£f 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 1ost consequent to a load drop. 2.2 C'O!l!"AJINMENT LOAD HANDLING SYSTEMS 2.2.l Jr.oad Handling Systems Capable of Carrying a Heavy Load (as defined in BU&l!tG-0612) in the Vicinity of the Reactor Vessel 2.2.1.1 Smnmary of Licensee Statements and Conclusions

• '!be fellowing load handling systems are capable of carrying heavy loads withilm 11:he containment building:

a poll.ar crane a aD!Wlus monorail o jib cranes. 2.2.1.2 ~~ation and Conclusion

'mine llacensee's identification of load handling systems within containment that are 1tc be evaluated for compliance with NUREG-0612 is consistent with the objectives of that document.

2.2.2 Beactor Containment Polar Crane 2.2.2.1 Summnary of Licensee Statements and Conclusions (lpeJrall:i.on of the reactor containment cranes is administratively contr@JU.ed.by Surry Technical Specification 3.10, which prevents handling of heavy JI.Jaaiils over the reactor vessel when there is fuel in the vessel unless the ]jift is specifically required for refueling operations.

We$tinghouse has been caaibracted to anaiyze the effects of load drops onto the vessel. The resll11ts

<mi: this analysis will be provided when available.

The Licensee does not imtcemill

~-o analyze the drop of heavy loads into the reactor vessel * * * ~klin Research Center M.!l!liision ol The Franklin lnslilule e TER-C5506-508/509 2.2.2.2 !lr.11.uation Ait:hough administrative controls may indicate that it is unlikely that a load dEap vil.l damage fuel in the reactor vessel, insufficient information has been provided to evaluate the individual loads that are lifted and the specific PEocedures followed during refueling operations.

Administrative controls may be found to support a degree of reliability consistent with the objectives of NUREG-0612 if it can be demonstrated that substantial margins of safety exist between the administratively controlled limit and the point at which daail!Je may occur. Based on the information currently provided by the Licensee, it cannot be concluded that administrative controls alone will reduce tine likelihood of a load drop from the polar crane in a manner that will prC17ide defense-in-depth load handling reliability and safety that is consistemit vith the guidelines of NUREG-0612, Phase II, Section 5.1.3. '!he Licensee has not provided sufficient information to evaluate the effects cf load drops*onto the reactor vessel. Although the Licensee has stated tbat Westinghouse is performing an analysis of. the-effects of load drops oimtc the reactor vessel, a determination must be deferred until the anal.ysis is C'Clllplete and can be compared with evaluation Criteria I through III of 1llilll!iR!lG-012, Section 5.1. I!msmfjficient information has been provided by the Licensee to ~tt:Jl.y determine compliance with NUREG-0612, Section 5.1.3 for the reactor CICl!ltainment polar crane. As a minimum, the Licensee should provide the fc.JLll.oJfilm!I information to satisfy the guidelines of NUREG-0612, Section 5 * .1.3: o J4iemtify the heavy loads that may be carried in the vicinity of the reactor vessel. c Jra!emtify specific procedures and procedural controls to be followed allni!!ml

~eavy loads are lifted in the vicinity of the reac~or vessel and ..fimelL is in the vessel. o PJr\OIVide the evaluation of the effects of load drops onto the reactor wessel when fuel is stored in the vessel and evaluate these results vitt!IR respect to the evaluation criteria of NUREG-0612, Section 5.1 * * * ~n Research Center .l!ll!lllliilian ol The Franklin lnslilute -s-2.2.3 Reactor Containment Annulus Monorails and Jib Cranes 2.2.3.1 Summary of Licensee Statements and Conclusions TER-CS506-508/509 The reactor containment annulus monorails and jib cranes are incapable of carrying heavy loads either directly over the reactor vessel or to a location where, in the event of a load handling accident, the load may land in or on the vessel .. 2.2.3.2 Evaluation flle safe load path sketches provided in Reference 5 indicate that it is highly unlikely that these cranes could carry a load to a location where, in the event of an accidental load drop, the vessel could be damaged *

• 2.2.3.3 Conclusion flle information provided in Reference 5 demonstrates that the reactor vesse1 containment annulus monorails and jib cranes meet the guidelines of NUREG-@612, Section 5.1.3. 2.3 1"1JEL BUI~DING OVERHEAD HANDLING SYSTEMS 2.3.1 Load Handling Systems Capable of Carrying a Heavy Load (as defined in HUREG-0612) in the Fuel Building 2.3.1.1 Summary of Licensee Statements and Conclusions 21me following load handling systems are capable of carrying heavy loads withiml the fuel building:

c new fuel crane c motor-driven platform and hoists fuel building trolley. 2.3.1.2 Eval.uation and Conclusion

'mine Licensee's statement and conclusions with respect to tbe load systems mentioned above are consistent with the information evaluated in Reference 4 and with the intent of NUREG-0612

• * * ~nkJin Resean:h Center A DMsiort cl The Fm:rilli:,i e

• TER-C5506-508/509 2.3.2 New Fuel Crane 2.3.2.l Summary of Licensee Statements and Conclusions The new .fuel crane is physically incapable of moving its hook centerline closer than 15 ft from the spent fuel pool boundary.

2.3.2.2 Evaluation NUREG-0612, Section 5.1.2 provides guidelines concerning the design and operation of load handling systems operating in the vicinity of the spent fuel pool. Safe load path sketches provided by the Licensee indicate that the new fuel crane is physically incapable of carrying heavy loads in the vicinity of the spent fuel pool. In addition, this information indicates that an accident load drop from this crane.could not affect the spent fuel pool. 2.3.2.3 Conclusion The operation of the new fuel crane at Surry Units 1 and 2 ls consistent with the staff's objectives for the safe handling of heavy loads at nuclear power plants. The Licensee has satisfied the applicable guidelines of Section 5.1 of NUREG-Q612.

2.3.3 Motor-Driven Platform and Hoists 2.3.3.1 Summary of Licensee Statements and Conclusions The motor-driven platform spans the spent fuel pit and may be maneuvered over any part of the fuel building area. However, administrative controls restrict the movement of heavy loads over spent fuel. Surry Technical Specification 3.10 prohibits the 1110vement of heavy loads exceeding 110% of the w@ight of a fuel ass~mbly {not including the fuel handling tool) over spent fuel. The onl~.*heavy loads* handled by the motor-driven platform and hoists are the transfer canal doors, and their movement is administratively controlled by Surry Technical Specification 3.10 and Surry Operating Procedure 4.18. ~e use of these administrative controls assures that the consequences of a load drop satisfy the acceptance criteria of NOREG-0612

• * * ~nklin Research Center A Oivisian o/ The Franlrlln lnslilUte TER-C5506-508/509 2.3.3.2 Bva1uation Surry Technical Specification 3.10 and Surry Operating Procedure 4.18 prohibit the movement of the transfer canal door over spent fuel~ however, prior to refueling, the transfer canal door is carried over the southernmost row of fuel storage positions.

In addition, the spent fuel positions will be used in such an order that the southerrunost row will be filled last [7]. Consequently, the Licensee has implied that the southernmost spent fuel rack will either reaain empty or the transfer canal door will not be moved over the rack if spent fuel is stored there. However, even when the southernmost spent fuel rack remains empty, the load path is still adjacent to other racks where spent fuel is being* stored. Therefore, the Licensee's reliance on adminstrative controls does not satisfy the Phase II requirements of NUREG-0612.

Additional information should be provided to demonstrate that sufficient separation exists between the load path of the canal gates and the stored spent fuel to preclude the possibility of a load drop damaging spent fuel or that the effects of a load .drop of the canal .gate will satify the evaluation criteria of NUREG-0612, Section 5.1. 2.3.3.3 Conclusion The Licensee has not demonstrated that administrative controls will satisfy the guidelines of NUREG-0612, Section s.1.2. Additional information should be provided to demonstrate that sufficient separation exists between the trave1 path of the canal gate and spent fuel, or the effects of the load drop of the canal gate should be analyzed to determine if they are consistent with the guidelines of NUREG-0612, Section s.1.2. 2.3.4 Fue1 Bu~lding Trolley 2.3.4.+ SIDlllilrY of Licensee Statements and Conclusions

• The fue1 buil.ding trolley does not carry any heavy loads, including the spent fuel cask, over any stored spent fuel. This trolley moves in a south direction over the east end of the fuel pool, where a built-up pad of * * ~ranidin Research Center A IDlioisimo d 1l!ie Frenlclin Institute

/ /

TER-C5506-508/509 energy-absorbing materials is located in the bottom of the fuel pit in the area where the spent fuel cask is loaded. This built-up pad was designed to reduce the consequences of a spent fuel cask drop so that structural damage to the reinforced concrete fuel pit structure, including cracking, will not occur. The consequences of a fuel cask dropped into the fuel pool were addressed in the Surry Final Safety Analysis Report as part of the original plant design and licensing.

Surry Technical Specification 3.10 prohibits the movement of a spent fuel cask into the fuel building until the NRC has ed and approved the spent fuel cask drop evaluation.

Upon resolution of this issue, compliance with NUREG-0612 will be established.

2.3.4.2 Evaluation To allow the movement of the spent fuel shipping cask into the fuel building, the Licensee requested a change to Surry Technical Specification 3.10 [BJ and provided supplemental information to this request in Reference

9. The NRC issued Amendment 84 to Facility Operating License DPR-32 and Amendment 85 to Facility Operating License DPR-37, (10] approving these requests.

In Reference 8, the Licensee provided an analysis of the effects of load drops of the spent fuel shipping cask onto the spent fuel pool. This analysis addressed the guidelines of NUREG-0612, Section 5.1.2(3) and evaluated the effects of the load drops with respect to the evaluation Criteria I through III of NUREG-0612, Section 5.1. Two casks were considered in the analysis to envelope the range of casks that potentially could be used ~t the Surry plant. The load handling accidents were evaluated under worst-case conditions consistent with the guidelines contained in Appendix A of NUREG-0612.

In Reference 9, the Licensee indicated that the spent fuel pool will be divided into a two-region storage pool; with Region 1 comprising the first three rows of fuel racks adjacent to the trolley l:9ad block. Region 2 will encompass the remainder of the fuel racks in the pooi. Region 1 will be limited.'to storage of spent fuel assemblies which have decayed at least 150 days. I ~nklin. Research Center A OMsion d/111,e Fmnldin Institute

• • TER-C5506-508/509 The following paragraphs are a brief summary of the Licensee's compliance with Criteria I through III of NUREG-0612, Section 5.1. Criterion I Following the postulated drop of the cask into the spent fuel pool, the Licensee assumed that the gap activity for the fuel assemblies stored in the first three rows of the spent fuel racks is released.

This is based on the NUREG-0612 guidelines that no *hot* spent fuel is to be stored within 25 ft of the trolley load block. The results indicate that offsite doses from the postulated accident will be maintained well below the limits of lOCFRlOO [11] by ensuring that all new fuel and spent fuel which has not decayed for at least 150 days will be excluded from the first three rows of the spen,t fuel racks. The first three rbws of racks encompass an area out to 28 ft from the load block. The results of this analysis satisfy Criterion I of NUREG-0612, Section 5.1. Criterion II The criticality analysis was performed for spent fuel stored in the first three rows of *the spent fuel racks. A matrix of criticality calculations was formed to determine the limiting fuel parameters that would allow fuel to be stored in *fuel racks, subject to damage from a cask drop accident without resulting in criticality.

In the analysis, the assembly-to-assembly spacing was reduced to maximize Keff for various combinations of initial enrichment and fuel exposure.

A graph (Figure 6.1 of Reference

8) was derived from these calculations, and the combination of exposure and initial enrichment assures that the maximum obtainable Keff occurs when the assembly-to-assembly spac-ing in the rack is reduced to approximately 6.9 in. The result.~ of the analysis indicate that spent fuel, which satisfies the initial enrichment vs. burnup limits in the graph (Figure 6.1 of Reference 8), can be safely stored in the first three rows of spent fuel racki which are subject.'to damage from a cask drop accident.

This will ensure that an accidental dropping of the spent fuel cask will not result in a configuration

• * ~nklin Research Center A Division ol The Franlclin lnllilule

• TER-C5506-508/509 of the fuel such that Keff is larger than 0.95, which satisfies Criterion II of NUREG-0612, Section 5.1. Criterion III The analysis evaluated the effects of postulated cask drops onto the fuel pool structure.

The accident was postulated to occur at any point along the cask travel path. The most likely point was above the cask loading area; both an edge drop and a flat drop of the cask were assumed. However, two cask impact pads protect the pool floor in the cask loading pit. These impact pads absorb most of the impact energy and limit structure damage to the pool floor. The structural analysis of damage to the fuel pool floor includes the mitigating effects of the cask impact pads. The results of the analt9is # indicated that a shear plug in the floor is not anticipated for a flat drop, concrete scabbing is not expected for an edge drop, and the maximum concrete _indentation ove~ .the impact area is not expected to tear or perforate the steel liner. The analysis of damage to the spent fuel pool walls assumed that all energy from the cask hitting the bottom of the pool was transmitted to the wall as rotational energy, i.e., no energy is lost at floor impact due to the cask pads' interaction.

The cask was assumed to drop to the bottom of the pool at the orientation that would cause the greatest angular acceleration into the wall. The results of this analysis indicate that scabbing would not occur, but penetration into the wall is possible.

This would cause pool liner damage, but no significant leakage is predicted and sufficient borated makeup water is available to assure that the fuel remains covered. Therefore, the analysis satisfies Criterion III of NUREG-0612, Section s.1. 2.3.4.3 Conclusion B~sed upon the NRC's approval of the Licensee's request, i.e., Amendments

• 84 and 85 [101, the operation of the fuel building trolley is consistent with . the guidelines of NUREG-0612, Section 5.1.2(3).

The Amendments revise Technical Specification 3.10 by eliminating the restriction on the handling of * ~nklin Research Center A Division ol The Franklin Institute TER-C5506-508/509 spent fuel casks in the spent fuel building.

The load drop analysis included in the Amendments satisfies evaluation Criteria I through III of NUREG-0612, Section 5.1. 2.4 OVERHEAD HANDLING SYSTF.MS IN AREAS CONTAINING SAFE SHUTDOWN EQUIPMENT 2.4.1 Load Handling Systems Capable of Carrying a Heavy Load (as defined in NUREG-0612) in the Vicinity of Safety-Related Equipment 2.4.1.1 Summary of Licensee Statements and Conclusions The following load handling systems are capable of carrying heavy loads in the vicinity of safety-related equipment:

o reactor containment polar crane o reactor containmept annulus monorails o reactor containment jib cranes o new fuel crane o fuel building trolley o decontamination building crane o 10-ton monorail system o 6-ton monorail system o filter cartridge removal monorail o Unit 1 switchgear room monorail o emergency diesel generator room monorails.

2.4.1.2 Evaluation and Conclusion The Licensee's statements and conclusions with respect to the load handling systems mentioned above are consistent with the information evaluated in Reference 4 and with the intent of NUREG-0612.

2.4.2 *Reactor Containment Polar Crane 2.4.2.1 Summary of Licensee Statements and Conclusions A load drop analysis for the reactor containment polar crane was completed and is provided in Table 1 of Reference

5. This analysis investigated the resul.ts of postulated drops of major loads cart"ied by the polar ctane onto the refueling cavity structure, reactor coolant pressure boundary, and various safe shutdown and decay heat removal systems. In the * ~nklin Research Center A OMsion al The Franklin lnslilUle e TER-C5506-508/509 case of damage to the refueling cavity structure or reactor coolant pressure boundary, the Licensee's analysis indicated that such an event would not prevent decay heat removal by alternate means. In the case of potential damage to safe shutdown and decay heat removal systems, the Licensee's analysis indicated that system redundancy and physical separation are sufficient to ensure that a loss of system function will not occur as a result of the load drop. The potential drop of the polar crane load block was eliminated from detailed evaluation since the testing of block limit switches will be administratively controlled and performed in a location beneath which there is no safety-related equipment.

2.4.2.2 Evaluation The load/impact area combinations identified and evaluated by the Licensee demonstrate that the intent of NOREG-0612, Section 5.1.5 has been satisfied through system redundancy and alternate means of heat removal. In the case of the octagonal floor plug, the Licensee has not provided sufficient information to independently determine whether there is any safe . shutdown or decay heat removal equipment located below the load path. 2.4.2.3 Conclusion The Licensee has partially satisfied the guidelines of NUREG-0612, Section 5.1.5 for the reactor containment polar crane. Additional information should be provided for the handling of the octagonal floor plugs, either through a site visit or additional drawings with a level of detail sufficient for an independent evaluation that all reasonable load paths and related components have been considered in the Licensee's analysis.

2.4.3 Reactor Containment Annulus Monorails

• 2.4.3.l . Summary of Licensee Statements and Conclusions A load drop analysis was completed and *is provided in Table l of Reference

5. This analysis investigated the results of postulated drops of ~nklin Research Center A Division ol The Franldln JnSlitule

• e TER-C5506-508/509 undefined loads weighing up to 5 tons and carried by the reactor containment annulus monorail.

The analysis indicated that a postulated load drop could affect safety-related equipment in the annulus area. However, the related equipment in the annulus area has adequate system redundancy and separation to preclude the loss of any system's safety function.

2.4.3.2 Evaluation The load/impact area combinations identified and evaluated by the Licensee, although reasonable, cannot be independently evaluated on the basis of the information provided by the Licensee.

2.4.3.3 Conclusion Additional information should be provided, either through a site visit or additional drawings with a level of detail sufficient for an independent evaluation, to verify that the safety-related systems in the annulus area have adequate system redundancy and separation to preclude the loss of system function following a load drop. 2.4.4 Reactor Containment .Jib Cranes 2.4.4.1 SU111Dary of Licensee Statements and Conclusions A load drop analysis has been completed and provided in Table l of Reference S. This analysis investigated the results of postulated drops of undefined loads weighing up to 8 tons and carried by the jib cranes. The analysis indicated that a postulated load drop could affect the reactor coolant pressure boundary.

However, system redundancy and separation preclude the loss of reactor pressure.

In addition, load drops that damage the reactor coolant pressure boundary do not prevent decay heat removal by the alternate systems availab_:t.e for operation if required.

• 2.4.4.2 Evaluation The load/impact area combination identified and evaluated by the Licensee, a1though reasonable, cannot be independently evaluated on the basis of the information provided by the Licensee.

• * ~ranidin Research Center A Diiiaia,, ot The Franklin Institute

• TER-C5506-508/509 2.4.4.3 Conciusion Additional information shou1d be provided, either through a site visit or additional drawings with a level of detail sufficient for an independent evaluation, to verify that the reactor. coolant pressure boundary has adequate system redundancy and separation and that the alternate systems are available for decay heat removal. 2.4.5 New Fuel Crane 2.4.5.l Summary of Licensee Statements and Conclusions A load drop analysis was caapleted and is provided in Table 1 of ence 5. This analysis investigated the results of postulated drops of the fuel container (full) and.the reaovab].e slabs onto the spent fuel pi~ pumps. *motor control centers. The new fue1 contaiqer (full) was also postulated to impact the spent fuel pit pumps and fuel pit cooler. However, for all of these postulated load drop accidents, the Licensee indicated that continued decay hea~ removal can be provided through the use of temporary makeup supplies of unborated water. Repairs can then be made while this unborated water is supplied.

In addition, the fuel storage rack configuration is such that the spent" fuel array will remain subcriticial in an unborated pool. 2.4.5.2 Evaluation The load/impact area combination identified and evaluated by the Licensee, although reasonable, ccaumot be independently evaluated on the basis of the information provided by tl!De JL.icensee.

2.4.5.3 Conclusion Additional information shoatld be provided, either through a site visit or additional draw'ings with a level. cf detail sufficient for an independent evaluation, to verify that tempor4llY makeup supplies of unborat~d water are availabl,e and that the spent fue]. pit pumps and fuel pit cooler have adequate system redundancy and separatiOl!ll to JP)reclude the loss of system function.

• I * ~nklin ,R:search Center A Division oi The Franklln lnslitute

_;_ TER-C5506-508/509 2.4.6 Fuel Building Trolley 2.4.6.1 Summary of Licensee Statements and Conclusions A load drop analysis was completed and is provided in Table 1 of Reference

5. This analysis investigated the results of postulated drops of the fuel _shipping cask, the bottom block and hook, and the spent resin shipping _container and cask. The Licensee's analysis indicated that the spent fuel shipping cask could drop onto the fuel pit cooling system piping. However, site-specific considerations eliminate the need to consider this accident.

In addition, the spent fuel storage rack is designed to remain subcritical under unborated water conditions.

Temporary makeup to the spent fuel pool using unborated water will maintain spent fuel cooling and shielding while repairs are made.

• For the bottom block and hook, the Licensee's analysis indicated that no safe shutdown or decay heat removal equipment is located below the limit switch testing l?Cation.

For the spent resin shipping container and cask, the Licensee's analysis indicated that no safe shutdown or decay heat removal equipment is located under the load path. 2.4.6.2 Evaluation In the NRC's approval of the Licensee's request to change Surry Technical Specification 3.10 [10], the Licensee also analyzed the effects of the spent fuel shipping cask being dropped onto the fuel pool cooling system piping in the pool and pipe trench at the northeast corner of the pool. The analysis indicated that damage to this piping would not cause the pool to drain and temporary repairs cou1d be affected, if necessary, to restore the piping system. A backup water supply from the fire hose stations is also available for cooling and shielding of the fuel while repairs are made to any damaged *. piping: * . Th~ transfer of the spent fuel shipping cask in the decontamination building was also eva1Wlted in Amendments 84 and 85 (10). Although no safe * ~nklin Researohi Cell'llter A ID!'vision of The F....Wml~

• TER-C5506-508/509 shutdown equipment is located in this building, there are two potential sources of radioactive releases beneath the cask travel path. These potential sources are the fluid waste treatment tank and the spent resin dewatering tank. However, the results of the analysis indicated that a cask drop onto these tanks resulting in the full release of the contents of each tank to unrestricted areas is not in excess of the maximum permissible concentrations (MPC) of 10CFR20, Appendix B, Table II, Column 2 [11]. Therefore, the consequences of a postulated load drop by the fuel building trolley while operating in the decontamination building are within acceptable limits. The evaluation of the bottom block and hook and the spent resin shipping container and cask, although reasonable, cannot be independently evaluated on the basis of the information provided by the Licensee for movement i~ the fuel building.

2.4.6.3 Conclusion The information provided in Amendments 84*and 85 [10] for the spent fuel shipping cask satisfies Criterion IV of NOREG-0612, Section 5.1. Additional information should be provided, either through a site visit or the provision of additional

'drawings with a level of detail sufficient for an independent evaluation, to verify the following:

o no safe shutdown or decay heat removal equipment is located below the limit switch testing location for the bottom block and hook o no safe shutdown or decay heat removal equipment is located under the load path of the spent resin shipping container and cask. 2.4.7 Decontamination Building Crane 2.4.7.l SUDllary of Licensee Statements and Conclusions A load drop analysis was coapleted and is provided in Table 1 of Reference

5. ~is analysis investigated the results of postulated drops of undefin~ loads weighing 5 tons and carried by the decontamination building crane. The Licensee's analysis indicated t~at a postulated load drop may fall and strike a spent fuel shipping container due to load swing. However,.

the * ~ranldin Research Center A DiilBian ol The Franldin illllsliblle e

• TER-C5506-508/509 spent f~ slln:ii.pping container is closed prior to its movement into the decent~ buil.ding so that any possible spillage of its contents is only a remote pass:ii.bi1ity.

The Licensee's analysis also indicated that no safe shutdow.m

(!J)JC' aat:ay beat removal equipment is located under the load path of this crame. 2.4.7.2 ~tion The aDDiiall.)'Sis of the spent fuel shipping container being tipped over is consistem1t vi.1th the intent of NUREG-0612.

However, the Licensee's analysis also i~tt:ed! that no safe shutdown or decay heat removal equipment is located tlDIDilll!!lr tile 1oad path of this crane. Although reasonable, the Licensee has not~ sufficient information to independently determine wb1!ther , this statt:e!ml!!lm¢ is accurate.

A&til:ii.¢ilfflJMI]

information should be provided, either through a site visit or additiamaall.

~s with a level of detail sufficient for an independent evaluatimm.

1t4l> ,zerify that there is no safe shutdown or decay heat removal equipmemtt: .llocatt:efil Wider the load path. 2.4.8.1 Sammait]'

cf Licensee Statements and Conclusions A :ll.maiB ~*anal.ysis was completed and is provided in Table l of Referem:&2 Si. 'lflmu. anal.ysis investigated the results of postulated drops of removablle!

sllabs all1ld IIUldefined loads onto the seal water heat exchanger, . nonregems:att:.i'R!

beat exchanger, the chemical and volume control piping, and the chillJleal

~nent cooling.exchanger and piping. The load drop analysis indicat:eml tt:lhla¢ 1t!llre 1oss of the seal water heat exchanger and nonregenerative a loss of capability for safe shutdown and

• decay~~~ because reactor coolant pump (RCP) operation can continue indefimlii.t:telly

~tlhotmt seal injection flow if component cooling water (CCW) is availcdrule.

e e TER-CS506-508/509 The ammal.ysis also indicates that system redundancy and separation will preclude tine l.c5s: of the chemical and volume control piping or the chilled component excbamY!fer and piping. Load!-carnyiJlg height restrictions are also followed during the movement of the umi!efinedi loads. A 1-foot maximum load carrying height restriction will be impn:sed.

on all lifting operations in the auxiliary building on elevations 27/ ft 6 in and 13 ft O in and will be administratively controlled.

2. 4. 8. 2 itvaJLUlliltiOD A review of the COi system at Surry Units 1 and 2 indicates that damage to eitheJ:' tine seai1 water or nonregenerative heat exchanger may result in damage tc 1tlme CCVI system supply to these heat exchangers, which may qot be isolab1e fr(Clll tine supply to the RCP thermal barriers.

The Licensee's system anal.ysis for this: accident sequence should be clarified.

Al~ imr>>t. strictly complying with the alternatives provided in HUREG-06JL2,, it :iis recognized that under certain circumstances admini*strative controls camm be employed to demonstrate that it is unlikely that a load drop wi11 safetr-rel.ated equipment.

Such circumstances will include substantial1 lllil.lC$:iilms between the administratively controlled limit and the point at .tnidm <iinrage aay occur and a high degree of certainty that any vic.ll.ati<<J1111 of the administratively controlled boundary can be promptly detected and actel! on !by smpervisory personnel.

Evidence of such circumstances has not been pireseimtted tile Licensee.

'!line LiceimSM sboul.d provide additional information to clarify the system amaJl.ysis of tt:he <<:Ci system following a loss of the seal water or tiwe lhleatt:.

~-'!llne LiccenSB!'~S a)PPa%ent reliance on administrative control~ to impose a load-cau:qiim!J lme:ii.9flnt restriction in the auxiliary building in the vicinity of safeity-fre.llalited elljfllli.paent has not been supported by information sufficient to *

• l&em:iimdb, ~ter At.l!iliiisimlllSl1lltl!

lmm!lllmlhdtule e TER-C5506-508/509 conclude 11:lmat the approach provides a degree of reliability consistent with the objectives of NUREG-0612.

2.4.9 Mionorail System 2.4.9.1 Smmra~ of Licensee Statements and Conclusions A load drop analysis was completed and is provided in Table l of Reference

5. '!'his analysis investigated the results of, postulated drops of the compameimt cooling water pumps and pump motor, the charging pump and pump 11K>tor, time removable slabs, and an undefined load from the 6-ton monorail.

The COii p!Jl!Dp or motor and the charging pump or motor are postulated to drop onto tile component cooling piping. The charging pump and pump motor are both posftlm].ated to drop opto the chemical and volume control piping and the component coo1ing piping. A drop of the removable slabs or the *undefined load* is postulated to drop onto the Unit 1 or 2 charging pump discharge headers piping, the Unit l or 2 safety injection piping, or the charging pump seal cocler surge tank. The analysis of all of these load drops indicated that system redundancy and separa,tion would preclude the loss of capability of the system to perform its safe shutdown,or decay heat removal function.

The ~emovable slabs and the undefined load could also drop onto the service V!l'alter piping to the charging pump intermediate seal coolers or the componemit coo1ing piping to the fuel pit coolers. For these load drops, system redundancy and separation preclude the loss of capability of the system to perf010a its safe shutdown or decay heat removal function.

In addition, the Licensee indicated that it can be shown by analysis that load drops will not damage Afe shutdown or decay heat removal equipment.

A load-carrying height restriction will also be followed which is administratively controlled.

A 1-foot maxia:u:a load-carrying height restriction will be imposed on all of the li£timJ operations in the auxiliary building on elevations 27 ft 6 in and 13 ft O up .. * ~nklin Research Center A Diwiliml oflhe Franklin lnslilute

• TER-C5506-508/509 2.4.9.2 Evaluation The load/impact area combinations identified and evaluated by the Licensee, although reasonable, cannot be independently evaluated on the basis of the information provided by the Licensee.

Although not strictly complying with the alternatives provided in NUREG-0612, it is recognized that under certain circumstances administrative controls can be employed to demonstrate that is unlikely that a load drop will damage safety-related equipment.

Such circumstances include substantial margins between the administratively controlled limit and the poin; at which damage may occur and a high degree of certainty that any violation of the administratively controlled boundary can be promptly detected and acted on by supervisory personnel.

Evidence of such circumstances has not been provided by the Licensee.

2.4.9.3 Conclusion Additional information should be provided, either through a site visit or additional drawings with a level of detail sufficient for an independent evaluation, to verify that system redundancy and separation will preclude the loss of safe shutdown or decay heat removal systems. The Licensee's apparent reliance on administrative controls to impose a load-carrying height restriction in the vicinity of safety-related equipment bas not been supported by information sufficient to conclude that this approach provides a degree of reliability consistent with the objectives of NUREG-0612.

2.4.10 Filter Cartridge Removal Monorail 2.4.10.l Summary of Licensee Statements and Conclusions A load drop analysis was completed and is provided in Table 1 of Reference

5. This analysis investigated tne cesults of postulaled drops of an undefined load weighing 2 tons * * ~nklin Research Center A OiviSIOII ol The Fl'llllklin lnslitule

• r I

• e TER-C5506-508/509 The analysis evaluated the postulated load drop of the undefined load onto the RCP seal water injection filter and associated piping. The analysis indicated that, due to system redundancy and separation, the loss of bility of the system to perform its function is precluded and the loss of all seal wate~ injection flow will not affect the operability of the reactor coolant pumps if component cooling water is available.

Unidentified specific considerations are also relied upon by the Licensee to eliminate the need to consider certain load and equipment combinations.

2.4.10.2 Evaluation The load/impact area combination identified and evaluated by the Licensee, although reasonable, cannot be independently evaluated on the basis , of the information provided by the Licensee.

In addition, the Licensee has not identified the site-s~ecific considerations upon which it is relying. 2.4.10.3 Conclusion Additional information should be provided, either through a site visit or additional drawings with a level of detail sufficient for an independent evaluation, to verify that adequate system redundancy and separation exist to preclude the loss of system function.

The Licensee should also identify the site-specific considerations which are relied upon in the analysis.

2.4.ll Unit l Switchgear Room Monorail 2.4.11.l Summary of Licensee Statements and Conclusions A load drop analysis was been completed and is provided in Table 1 of Reference

5. This anaJ.ysis investigated the results of postulated drops of the motor generator set motor onto miscellaneous cables and conduits and the control room roof. The analysis indicated that a load drop onto some misceflapeous cables and con4uits would be resolved through system redundancy and-separation.

and a drop of the motor generator set motor onto ~he control room roof would not * * ~nklin Research Center A Oivision ol The Franklin Institute e e TER-C5506-508/509 affect any safety-related equipment.

The control room roof is designed to withstand a tornado-generated missile. The kinetic energy developed by a drop of the motor generator set motor is significantly less than the kinetic energy developed by a tornado-generated missile. Therefore, the-control room roof will not be damaged by a drop of the motor generator set motor. 2.4.11.2 Evaluation and Conclusion The evaluation of a load drop onto some miscellaneous cables and conduits demonstrated that system redundancy and separation preclude the loss of system function.

In addition, the load/impact area combination of the motor generator set motor onto the control room roof indicates that no damage to safety-related equipment can occur. The information provided in the load drop analysis demonstrates that the guidelines of NUREG-0612, Section 5.1.5 are satisfied.

2.4.12 Emergency Diesel Generator Room Monorails 2.4.12.l Summary and Licensee Statements and Conclusions A load dr.op analysis was completed and is provided in Table l of Reference

5. This analysis investigated the results of postulated drops of diesel generator parts weighing up to 10 tons onto an emergency diesel generator.

The analysis indicated that an accidental load drop from either of these monorails would affect only the emergency diesel generator being -repaired.

However, the availability of the other diesel generators which are located in separate diesel generator rooms precludes the loss of capability of this system to perform its safety-related function.

2.4.12.2 Evaluation and Conclusion The information provided by the Licensee demonstrates that the emergency diesel*generator room monorails satisfy the guidelines of NUREGr0612, Section 5.1. 5. , ~klin Research Center A Divmor, ol The Franklin Institute

• TER-C5506-508/509

3. CONCLUSION This summary is provided to consolidate the results of crane-specific evaluations presented in Section 2. It is not meant as a substitute for the specific 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 addressed in seeking to resolve issues where the degree of load handling reliability provided by cranes at the Surry Power Station 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 the guidance of NUREG-061}.

1. 3.1 INFORMATION ISSUES The following information provided by the Licensee has been assessed as insufficient to support an independent conclusion that load handling reliability is consistent with the evaluation criteria of Section 2.1 in the following areas: Load/Impact Area Data (Sections 2.4.2, 2.4.3, 2.4.4, 2.4.5, 2.4.7, and 2.4.10) The Licensee identified specific safety-related equipment and/or components that would be affected by a heavy load accidently dropped from a crane and the hazards elimination category relied upon to resolve each situation.

However, the plant layout drawings and safe load path diagrams do not provide sufficient information to independently determine whether the information provided is accurate.

Hazard Elimination Category (Sections 2.4.6 and 2.4.10) The Licensee identified "site~specific considerations" as a hazard elimination category in the load/impact area data. However, no details ot these site-specific considerations have been identified in the table, and therefore it cannot be determined whether these considerations are consistent with the intent of NUREG-0612.

~nklin Research Center A Om,,ia,, cl The Franklin Institute

• e TER-C5506-508/509 3.2 APPROACH ISSUES The approach or position taken by the Licensee, based on information provided thus far, appears inconsistent with the staff's objectives in the following respects.

Use of Administrative Controls (Sections 2.2.2, 2.3.3, 2.4.8, and 2.4.9) 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 irradiated fuel and safe shutdown equipment.

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 over spent fuel or safe shutdown equipment.

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 monitorable and strictly enforced), administrative controls ca~ be found, on the basis of engineering judg~ent, 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

• * ~nklin Research Center A Oh,ision ol The. Franklin Institute

• TER-CSS06-508/509

4. REFERENCES

l. 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, ncontrol of Heavy Loads at Nuclear Power Plantsn July 1980 3. 4. D. G. Eisenhut (NRC) Letter to All Operating Reactors

Subject:

Control of Heavy Loads December 22, 1980 FRC Technical Evaluation Report, *control of Heavy Loads at Surry Power Station Units land 2* TER-CSS06-395/396, January 14, 1983 5. R.H. Leasburg (VEPCO) Letter to D. G. Eisenhut (NRC)

Subject:

Control of Heavy Loads (Phase II) March 22, 1982 6. NRC NUREG-0544, *single-Failure-Proof Cranes at Nuclear Power Plants* May 1979 7. C. M. Stallings (VEPCO) Letter to V. Stello (NRC)

Subject:

Movement of Heavy Loads Near Spent Fuel October 9, 1978 8. w. L. Stewart (VEPCO) Letter to S. A. Varga (NRC) 9.

Subject:

Proposed Technical Specification Changes September 23, .1982 (NRC) W. L. Stewart (VEPCO) Letter to s*: A. Varga Su~ject: Supplemental January 17, 1983 Information for Proposed Operating License Amendaemt

~nklin Research Center A Division al The Fl?lnldin Institute

• *

~,-* 10. S. A. Varga (NRC) Letter to W. L. Stewart (VEPCO)

Subject:

Amendments 84 and 85 March 4, 1983 11. Code of Federal Regulations

-Energy 10-Parts Oto 199 January l, 1983 * ~nklin Research Center A Division of The Franklin Institute

• TER-CS506-508/509