ML20095H063
| ML20095H063 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 07/10/1984 |
| From: | Bomberger C, Sargent I, Walter R FRANKLIN INSTITUTE |
| To: | Singh A NRC |
| Shared Package | |
| ML20095H062 | List: |
| 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-52221, TAC-52271, TER-C5506-489, TER-C5506-489-DRFT, NUDOCS 8408280321 | |
| Download: ML20095H063 (26) | |
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TECHNICAL EVALUATION REPORT CONTROL OF HEAVY LOADS - PHASE ll TOLEDO EDISON COMPANY DAVIS-BESSE NUCLEAR POWER STATION UNIT 1 NRC DOCKET NO. 50-346 FRC PROJECT C5306 C
HRC1ACNO. 52271 FRC ASSIGNMENT 19 NRC CONTRACT NO. NRC-OH1 130 FRCTASK 489 Preparedby Franklin Research Center Author: C. R. Bomberger.
20th and Race Streets R. J. Walter Philadelphia, PA 19103 FRC Grouploader:
I. H. Sargent Prepared for Lead NRC Engineer:
A. Singh Nuct' ar Regulatory Commission e
Washington, D.C. 20555 July 10, 1984 This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Govemment 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-retus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights.
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TER-C5506-489 CONTENTS Section Title Page 1
INTRODUCTION 1
1.1 Purpose 1
1.2 Generic Background.
1 1.3 ' Plant-Specific Background.
2 2
EVALUATION.
3 2.1 Evaluation Criteria 3
2.2 Overhead Eandling Systems.
4 2.3 Spent Fuel Fool Area 5
2.4 Reactor Contairunent Area 9
2.5 Overhead Bandling Systems in Areas Containing Safe shutdown Equipment.
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COllCLUSIONS 21 4
REFERENCES 23 D
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I TER-C5506-489 FON l
This Technical Evaluation Report was prepared by Franklin Research Center under a contract with the U.S. Nuclear Begulatory Commission (Office of Nuclear Reactor Regulation, Division of Operating Reactors) for technical ccsistance 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, Mr. C. R. Bomberger, and Mr. R. J. Walter contributed to the technica.1 preparation of this report through a subcontract with WESTEC Services, Inc.
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1.
INTRODUCTIDN 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 rcactor shutdown and fuel element decay heat removal at Davis-Besse Nuclear Power Station Unit 1.
This review constitutes the second phase of a two-phase review instituted to resolve a generic issue pertaining to the safe handling cf heavy loads at nuclear power plants.
1.2
" w aC BACEGROUND Generic Technical Activity Task A'36 was established by the Nuclear Regulatory Comunission (NBC) 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 NBC staff on May 17, 1978 [1] to all power reactor licensees, requesting information concerniing the control of heavy loads near spent fuel.
The restalts of Task A-36 were reported in NUREG-0612 [2].
The staff concluded from this evaluation that existing measures to control the handling cf heavy loads at operating plants provide protection from certain potential 4
prsblems but do not adequately cover the major causes of load handling accidents and should be upgraded.
'Bo upgrade measures for the control of heavy loads, the staff developed a ccries 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 NUEEG-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 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 e QWCenter N
TER-C5506-489 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
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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 ccc:idents l'ndicate that the potential consequences of a load drop, are acceptably small.
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1.3 PLANT-SPECIFIC BACIGROUND On December 22, 1980, the NRC issued a letter [3] to the Toledo Edison Company (TEC), the Licensee for Davis-Besse Nuclear Power Station Unit 1, I
l r;:: questing 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 cf this independent evaluation with respect to general load handling equipment End procedures (Phase I) were provided on August 9, 19g3 [4].
On June 10,.
1983, TEC 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.
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2.
EVALUATION This section presents an evaluation of critical load handling areas at Davis-Besse Nuclear Power Station Unit 1.
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 alternative's is summarized and evaluated, and a conclusion as to'the extent of compliance, including recoussended additional action or requirements for additional information as appropriate, is provided.
2.1 EVALUATIDN CRITERIA 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 r'eactor 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 EUREG-0612, Sections 5.1.2, 5.1.3; and 5.1.5 (for pressurized troter reactors) and in 5.1.4 and 5.1.5 (for boiling water reactors) forms the basis for the conclusions reached in this siection and is briefly summarized as follows.
For a determination that the likelihood of damage is extremely smail:
o The design of the load handling system (i.e., crane or hoist and underhook lifting devices) is consistent with, or equivalent to, the NRC staff criteria for single-failure-proof cranes identified in NUREG-0554 [6), or 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 acceptables o In the case of potential damage to spent fuel, calculations have been provided to demonstrate that potential radiological doses at the site 4 11111111 p a.am. n
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l TER-C5506-489 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 E.gg larger than 0.95.
In the case of damage to the reactor vessel or spent fuel pool, it can o
be demonstrated that this damage will be limited to the extent that the fuel will not become uncovered.
In the case of damage to equipment or components employed for reactor o
shutdown or fuel element decay heat removal, it can be demonstrated that the safety-related function of the affected system will not be l
lost.
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~.1 2.2 DVERBEAD HANDLING SYST MS 2.2.1 Summary of Licensee Statements and Conclusions Informatica provided by the Licensee identified the following load handling systems, capable of carrying loads over the indicated areas, to be cubject to the Phase II criteria of NUREG-0612:
1.
In the vicinity of the spent fuel pool o spent fuel cask crane (140/20 tons) 2.
In the vicinity of the reactor vessel -
" containment polar crane (180/25 tons) l o
l o reactior service crane (5 tons) 3.
Over equipment required for safe shutdown o spent fuel cask crane o containment polar crane o component cooling water pump monorails o service water intake structure gantry crane.
2.2.2 Evaluation The identification of. the above load handling systems as being capable of carrying heavy loads within the reactor building is consistent with the intent cf EUREG-0612.
In addition, however, the containment equipment jib cranes were also identified in References 4 and 7 as load handling systems which are capable of
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TER-C5506-489 carrying heavy loads over the core, over spent fuel, or over safety-related equipment. The Licensee has not provided relevant information regarding these cranes in the Phase II' response [5] to NUREG-0612.
2.2.3 Conclusion The Licensee's statements and conclusions with respect to the load handling systems subject to the, criteria of NUREG-0612 are consistent with those identified in Beforence 5, with the exception of the containment equipment jib cranes. The containment equipment jib cranes were previously identified to be subject to the criteria of NUREG-0612; therefore, the Licensee should evaluate these cranes.for compliance with NUREG-0612, Phase II criteria.
2.3 SPENT FUEL POOL AREA 2.3.1 Spent Fuel Cask Crane 2.3.1.1 Summary of Licensee Statements and Conclusions The spent fuel cask crane is physically capable of carrying heavy loads over spent fuel in the spent fuel pool.
Plant Technical Specification 3.9.7 provides various physical and administrative controls to prevent loads greater than 2,430 lb (weight of one fuel assembly and its handling tool) from being carried over spent fuel in the pool.
The spent fuel cask crane is electrically interlocked to prevent the crane from traveling over the spent fuel pool while any load is suspended from the main book. The interlock can only be bypassed with a key obtained from the Shift Supervisor. Even upon bypassing, the main hook stays inoperative, cad only the auxiliary hook can be used.
The spent fuel pool divider gates (4 tons) are the only loads handled in l
the insediate vicinity of spent fuel in the spent fuel pool.
The Licensee analyzed the effects of an accidental load drop of the canal gates with rcepect to Criteria I through III of NUREG-0612, Section 5.1. 1 W EW
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Criterion I Assumptions used to determine the consequences of a load drop were based cn the Davis-Besse FSAR and Safety Guide 25.
Analyses performed by the Licensee indicated that the maximum number of fuel assemblies that could be damaged without exceeding NUREG-0612 guideline doses would be 15 assemblies.
For the most limiting drop geometry of the divider gates, results of the Licensee's load drop analysis indicate'that less than 15 fuel assemblies will be damaged,. and therefore, the Licensee concludes that MUREG-0612, Criterion I is satisfied.
criterion II Regarding the criticality analysis, the Licensee stated that the casumptions of NUREG-0'.12, Section 2.2 are consistent with the existing design at the Davis-Besse plant. The fuel spacing of the fuel storage racks is designed to maintain K,gg at a value of less than 0.90.
System Procedure SP i
1104.42, " Spent Fuel Pool Operating Procedure," requires that the boron concentration in the spent fuel pool be maintained at a value greater than or equal to 1800 ppe.
In addition, the License stated that the assumption was made that all fuel had an enrichment of 3.5 weight pere.ent, which is the
" highest probable enrichment" identified in the FSAR. Therefore, based upon a l
Davis-Besse design water-to-fuel ratio of 1.22, the Licensee stated that a l
load drop which causes crushing of the core will result in a decrease in I,gg and will not exceed 0.90, thereby satisfying Criterion II.
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Criterion III I
j To determine the possibility of a load drop cap. sing water leakage from the spent fuel pool, the Licensee performed a structural analysis of the potential for perforation or scabbing of the 5-ft spent fuel pool base.
Results of this analysis indicate that perforation and scabbing are not probable.
Penetration of the base is predicted to be insignificant and no
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leakage is anticipated, therefore satisfying this criterion.
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2.3.1.2 Evaluation Information provided by the Licensee identified two situations requiring evaluation of loads handled in the spent fuel pool area:
o unrestricted movements of loads over the spent fuel pool precluded by electrical interlocks o bypassing the electrical interlocks to allow movement of the pool divider gates by the spent fuel cask crane auxiliary book.
For routine movements of heavy loads outside.the spent fuel pool, use of clectrical interlocks to pr' event load movements over the pool satisfies, to a Icrge degree, the criteria of MUREG-0612, Section 5.1.2(2) for protection of spent fuel in the spent fuel pool.
Additional information is required, however, to verify that adequate physical separation exists between the limits cf crane travel (as restricted by electrical interlocks) and the walls of the cpent fuel pool so that a dropped load (i.e., spent fuel cask) cannot tip or rail and damage the opent fuel pool wall or roll into the spent fuel pool and damage spent fuel.
The Licensee identified the movement of the pool divider gates using the cuxiliary hook as the only instance in which the electrical interlocks will be bypassed. For this load movement, the Licensee performind analyses to demonstrate that the consequences of a load drop would not exceed Criteria I, II, and III of NU5tEG-0612, Section 5.1.
i criterion I The Licensee assumptions appear consistent with those of NUREG-0612, Appendix A, and the analysis results indicate that the resultant fuel damage vill not produce offsite radiological consequences which will exceed NUREG-0612 guidelines. Therefore, from information provided, it appears that criterion I is satisfied for a load drop in the spent fuel pool area bounded by the weight of the p.Jol divider gates.
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TEIH:5506-489 Criterion II Based upon comparison of the plant-specific design information provided by the Licensee with the information provided in Section 2.2 of NUREG-0612, it appears that K,gg following a load drop will not exceed a value of 0.90, which satisfies Criterion II.
This conclusion is based upon the following information provided by the Licensee:
o K,g.g of spent fuel is less than 0.90 based upon design. spacing of the spent fuel racks o spent fuel pool boron concentration is procedurally maintained greater than 1800 ppm o the " highest probable" fuel enrichment is 3.5 weight percent 4
o design water / fuel ratio is 1.22.
However, order to fully agree that these criteria will satisfy Criterion II criticality concerns, additional assurance should be provided to cubstantiate the following concerns.
Although the Licensee stated that plant procedures exist which require that boron concentration be maintained greater than 1800 ppm in the spent fuel pool, insufficient information has been provided to enture that the limit is cnforced on a continuing basis by periodic sampling or to identify corrective rctions to be taken in the event that concentration decreases to less than 1800 ppu. Similarly, although the Licensee assumes that all fuel impacted is I
cariched to less than 3.5 weight percent, additional information and assurances cre needed to ensure that these " highest probeble" FSAR values will not be CXceeded.
i Criterion III Information provided by the Licensee appears to demonstrate adequately that no leakage will occur from the spent fuel pool as a result of a load drop cf the pool divider gates, which satisfies this criterion of EUREG-0612.
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2.3.1.3 Conclusion Naasures implemented by the Licensee in the vicinity of the spent' fuel pool partially satisfy the criteria of NUREG-0612, Section 5.1.2(2).
To fully cctisfy the NUREG criteria, the following additional actions are required:
o For electrical interlocks which prevent load movement over the spent fuel pool, the Licensee should verify that adequate physical separation exists between the limits of crane travel and the spent fuel pool wall such that a dropped load will not impact, tip, or roll and cause damage to the spent fuel pool wall or spent fuel in the spent fuel pool.
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The Licensee should identify (1) the means of enforcing plant procedures on a continuing basis to ensure that boron concentration in the spent fuel pool does not decrease to less than 1800 ppa, as well as (2) the limitations imposed on load handling if this limit is violated.
It is also requested that the Licensee provide a more definitive statement to ensure that the " highest probable" enrichment will not exceed 3.5 weight percent.
2.3.2 Reactor Service Crane 2.3.2.1 Summary of Licensee Statements and Conclusions The reactor service crane is physically capable of carrying heavy loads over the re' actor. At this time, the crane is not in user therefore, the Licensee has deferred evaluation of this crane until it is got into service.
2.3.2.2 Evaluation and Conclusion The operation of the reactor service crane cannot be independently evaluated until the Licensee provides an analysis of this crane with respect to the Phase II requirements of NUREG-0612.
2.4 EEACTOR CONTAIIBEENT AREA 2.4.1 Containment Polar Crane 2.4.1.1 Susunary of Licensee Statements and Conclusions The containment polar crane is physically capable of carrying heavy loads over the reactor vessel. A load drop analysis for the postulated drops of the major loads carried by this crane is provided in Appendices A and B of
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TER-C5506-48 9 Reference 5.
The Licensee does not consider it feasible to postulate a random mechanical failure of the crane load-bearing components when moving either the main hoist or auxiliary hoist load block without a load and has not included this load in the load drop analysis. Bowever, the Licensee identified two possible failure modes that could result in the load drop of the main book and load blocks i
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A control system or operator error resulting in hoisting of the block to' a "two blocking" position with continued hoisting by the motor and subsequent parting of the rope (this situation can be prevented by operator action prior to "two blocking" or by an upper limit switch to terminate hoisting prior to "two blocking").
2.
Uncontrolled lowering of the. Iced block due to failure of the holding brake to function (the likelihood of this can be made small by use of redundant holding brakes).
To prevent the occurrence of these load drops, the polar crane main and cuxiliary hoists are provided with upper and lower limit switches along with a Revere digital weight indicator and limiter. The Bevere digital weight indicator and limiter significantly reduces the likelihood of damage to the crane or lifting devices due to an overload, and the limit switches reduce the likelihood of two blocking. The main and auxiliary hoists are also equipped with dual electric brake systems that reduce the likelihood of uncontrolled lowering.
Therefore, the Licensee concluded that a drop of the load block and hook 10 of sufficiently low likelihood that it does not require a load drop cnalysis.
To demonstrate compliance with the requirements of NUEEG-0612 for 3
remaining heavy loads, the Licensee has performed both structural and systems cnalyses to damaantrate that the consequences of a load drop will not exceed the criteria of NUREG-0612, Section 5.1.2.
In performing these evaluations, applicable heavy loads were identified, realistic load drop scenarios developed based on procedures in effect, and systems evaluations used to cugment structural analyses which indicated local failure only. The detailed ctructural analyses that were performed assessed the structural response and,
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' consequences of a dropped load. Structural evaluation methods and criteria fcilow the criteria of MUREG-0612 and the recommendations of the ASCE technical committee on' impulse and impact loads. Based upon conclusions contained in the following paragraphs, the Licensee concludes that NUREG-0612 Criteria I through III are met for all postulated load drops over the reactor vessel.
The structural evaluations were dividad into two categories:
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c using local failure only (spalling or scabbing) and those for which overall ctructural deformation or failure is anticipated. Eetailed evaluations were performed only for those drops whose consequences were unacceptable.
The Licensee's analysis included'an evaluation of a load drop of the plenum assembly (119,000 lb) into the core from the highest elevation possible (73.5 in), as dictated by physical restraints.
(S is height was determined by the height of the internals indes fixture, which is used to ensure exact clignment of the assembly as it is remoted and replaced.)
Se Licensee stated l
that the maximus kinetic energy associated with such a drop is 729,000 ft-lb, cssuming that the load reaches the core and impact is uniformly transmitted to cil fuel cells. The evaluatior. indicates that the strain in the individual fuel cells,beyond buckling does not exceed an allowable strain level of 0.01, based upon the properties of the cladding. Therefore,"it is the Licensee's conclusion that the fuel cladding will not rupture and no radioactive gases eill be released.
A load drop of the reactor vessel head (Rvu) has also been evaluat ed by the Licensee. The RVE (330,000 lb) has been postulated to drop from a height i
i of 5 f t (based upon procedural limitations), and the impact is transmitted j
from the RPV flange through the RFV shell to the cold leg nossles, which are I
the EPV ultimate supports. Evaluation indicates that the ultimate load that l
c:n be absorbed by thess supports is 7 million pounds, whereas the maximum kinetic energy that is produced by the load drop is only 1.65 million pounds and is therefore acceptable based upon accident assumptions.
The leak tight integrity of the reactor coolant pressure is also d==<==trated.
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An evaluation of a load drop of one of the six missile shields (94,500 lb) onto the control rod drive service structure has also been conducted. Procedures are in place which require the removal and replacement cf these shields in specific sequential order so as to minimize the potential danger to the components located below. The Licensee stated that these l
procedures ensure that a dropped shield will fall back into place or will fall onto another shield located below. Therefore, hosed upon these considerations cad the bounding load drop of the RVE, the consequences of a drop of a missile shield is acceptable.
In addition to the structural evaluations, the Licensee has also performed systems evaluations to demonstrate that acceptable consequences are present for specific cases.
Several load drop scenarios have been defined, based upon the status of the RVE (installed or removed) and the size of the resulting unisolable BCS leak. Imad drops reviewed in the reactor vessel area indicate that possible targets are the BCS mossles and the cere flood nossles.
Resulta j
cf the structural analysis indicate that the RCS nosales will remain intact.
Based upon the assumption that only one core flood nossle will rupture, the j
Licensee states that all cases that were evaluated indicate that adequate core l
cooling will be maintained.
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2.4.1.2 Evaluation
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The Licensee's analysis of the containment polar crane has been evaluated j
in accordance with the criteria of NUEEG-0612, Section 5.1.3, for loads handled in the vicinity of the reactor vessel. The Licensee has identified crane design features which aske the likelihood of damage from an unloaded load block entremely remote and has performed detailed analyses in an attempt to demonstrate that Criteria I, II, and III of NUEEG-0612 have been satisfied.
It is agreed that for an unloeded load block, suitable design features in the form of upper and lower limit switsbes, dual electric brake systems, use l
cf a weight indicator and limiter, and the large material safety margins associated with the lift of an unladen block eliminate the need for further analysis or crane modification.
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l TER-C5506-489 For other loads carried in the vicinity of the reactor vessel, the fc11owing are evaluations of the Licensee's statements with respect to the criteria of EUREG-0612', Section 5.1.
Criterion I Analysis of information provided by the Licensee appears to adequately demonstrate, that fuel will not be crushed and no adverse radiological consequences will be experienced, based upon the Licensee's conclosion that a drop of the 119,000-lh plenum assembly from a height of 73.5 inches is the limiting load drop condition. Bowever, it is requested that the Licensee confirm that no other situations exist in which a physically smaller load may be dropped from a higher height and cause sufficient local damage to exceed the offsite radiological ceasequences of Criterion I.
If the Licensee confirms that a drop of the plenum assembly bounds all other such load dropa, then Licensee assumptions are ocasistent with EUREG-0612 and the Licensee will satisfy criterion I in the vicinity of the reactor vessel.
Criterion II Drops of the following loads which may cause critihality conditions have been analysed by the Licensees plenum assembly gnd missile shields.
- Bowever, 00ither analysis appears to specifically address the issue of fuel crushing and the resultias ;:etantial for criticality. Therefore, the Licensee should provide additional information to confirm that these analyses demonstrate that criticality conditions will not be exceeded or perform appropriate analyses to address this issue.
Criterion III The Licensee has performed structural evaluations of two load drops in the vicinity of the reactor vessel, as well as a systems analysis of this arca, to demonstrate that criterion III will not be violated.
The two load drops are the drop of the reactor vessel head (EVE) from a height of 5 f t and a drop of a missile shield over the core.
Unlike the plenum assembly, a drop cf either load onto the reactor vessel does not require precision alignment E**
TER-C5506-489 t:ith an indexing fixture. The RVE drop height of 5 ft is based upon procedural limitations only, while the missile shield evaluation is also predicated on I
procedural controls in which the shields are moved laterally and sequentially cutward from the center of the reactor vessel at the minimum height necessary to clear other in-place shields. Orientation of the shields is not predicted by the Licensee to change, and therefore a drop will impact either another cissile shield or the D-ring wall.
Evaluation of initial conditions for both a RVE drop or missile shield drop indicates that both analyses rely on procedural restrictions which limit the consequences of the load drop and therefore do not conservatively predict worst-case consequencs associated with a load drop. These assumptions are not consistent with Appendix A of EUREG-0G2, which states that analyses should assume worst-case orientation and a drop from the =awimum height in deter-cining the consequences of a load drop. In general, such procedural controls cre not equivalent to physical restraint, enhanced load handling reliability, 1
or load drop analysis in demonstrating that loadM rop consequences are accept-cble or that the likelihood of a load drop is reduced. Therefore, structural l
cnalyses performed for the RVE and procedures controlling movements of the cissile shipids are not acceptable as a basis for ccepliance with Phase II.
In addJtion.to structural evaluations, the Licensee has performed systems analyses to verify that core cooling will not be lost. Although the Licensee indicates that core cooling will not be lost, the following assumptions should i
be verified to be accurate:
1 1.
The systems analysis is based upon results of the structural analysis in assuming that vessel integrity is not lost since it has been requested that the Licensee reconfirm the validity of structural evaluations for worst-case consequences, acceptability of the systems approach is contingent upon this reevaluation.
2.
No rationale has been provided to agree with the Licensee's assumption that only one of tuo core ficed nozzles will be ruptured.
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TER-C5506-489 2.4.1.3 Conclusions Although the Licensee has performed analyses to demonstrate compliarme tith the applicable criteria of NUREG-0612 for heavy loads handled in the vicinity of the reactor vessel, evaluation of these analyses indicates that ccsumptions used by the Licensee have not been clearly addressed or are not in Cccordance with the guidance of NUREG-0612, Appendix A.
Therefore, Licensee action is requested to address the remaining concerns for each of the ic11owing criteria identified as follows.
Criterion I The Licensee should verify that a drop of the plenum assembly, presents f
the bounding load drop for analyzing damage to the fuel and resultant off-site consequences.
Criterion II The Licensee should address the potential for exceeding limits for criticality specified in this criterion.
i criterion I'II The Licensee' should reanalyse the consequences of a load drop and its effects on structural integrity, using assumptions consistent with NUREG-0612, Appendix A, as well as the consequences of new structural analyses on the systems analysis.
Use of procedural controls alone is not an adequate alter '
n:tive to performance of analyses or improvement of load handling reliability to document conformance with Phase II guidelines.
2.5 OVIEEEAD NANDLING SYSTBIS IN AREAS CONTAINING SAFE SEUTDOWN EQUIPMElff 2.5.1 containment Polar Crane 2.5.1.1 Summary of Licensee statements and Conclusions The Licensee identified the containment polar crane to be capable of handling heavy loads over equipment required for safe shutdown.
To demon-
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6 TER-C5506-489 ctrate compliance with Criterion IV of NUREG-0612, Section 5.1, the Licensee performed detailed structural and systems analyses to demonstrate that core cooling can be maintained following any credible load drop.
The containment w s segmented into 9 regions for purposes of these analyses.
In each of these l
creas, the License performed specific structural analyses based upon the limiting load drop which may occur in the region and systems analyses based upon the safe shutdown systems present within the region.
Results of the integrated
- analyses are tabulated as follows:
Region 2 North End of Operating Deck The limiting load drop is a drop.of the reactor vessel head from a height j
cf 7 ft.
Structural analysis indicated structural deformations, and therefore c11 systems were assumed lost in the area.
However, systems analysis indicated that decay heat removal would remain available and core cooling could be maintained..
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Region 3:
D-Ring Enclosure The limiting load drop is drop of either a missile shield or a reactor l
coolant pump (Bri) motor into the steam generator area.' Due to the complexity of the structuzal analysis for a drop of the RCP motor, the analysis was not performed, and it was assumed that the decay heat removal system components t:ere lost. Bowever, the Licensee stated that the requisite emergency core cooling system (BCCS) components would remain available to maintain core cooling, or a suitable arrangement could be made by aligning the PORV system.
i Region 4 Refueline Canal The limiting load drop is a drop of the missile shield from a height of l
76 f t.
Because the consequences of a drop of the missile shields indicated cacessive structural deformations, a systems analysis was performed.
It was assumed that a LOCA would occur in one loop and that one train of core flooding would be damaged.
Results indicated that core cooling capabilities would not be lost and flooding and leakage level were acceptable with respect to survivability of equipment.
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TER-C5506-489 Region 5:
South End of Refueling Canal The limiting load drop is a drop of the plenum assembly on the cavity floor. Structural and systems analyses indicated that the consequences of such a drop would be insignificant and that load movements within this area cre acceptable.
Region 6 Piping Enclosures Impact due to secondary missiles. are the only potential problems in this crea because physical interference prevents movements of heavy loads over this area. Systems analysis assumed that one line of decay heat removal would be lost; however, analysis results indicated that adequate redundancy would exist to maintain core cooling.
Region 7 SE Quadrant Grating As in region 6, no heavy loads may be moved over this region. Assuming loss of decay heat removal suction piping in this area, analysis indicated that the ICC$ would still be available and core cooling would be maintained.
Region 8:
in-Core Instrument Area The limiting" load drop is a drop of the in-core instrument tank access h tch cover (3.4 tons). Structural analyria indicated' that a drop of this load will not perforate the 606-f t elevation slab.
Systems analysis assumed that all instrument tubes located in the trunk would be severed, causing a IDCA.
Analysis results indicated that the ECCS would still be available and core cooling would be maintained.
Region 9: Area Adjacent to Equipment Hatch The limiting load drop is a drop of a RCP motor. No structural analysis was performed. Systems anlysis assumed loss of makeup and purfication piping, r sulting in a IDCA. Analysis results indicated the ECCS would remain unaffected and core cooling would be maintained.
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TER-C5506-489 2.5.1.2 Evaluation The analyses performed by the Licensee to evaluate the consequences of load drops onto equipment required for safe shutdown appear to satisfy, to a 1crge degree, the evaluation criteria of NUREG-0612, Criterion IV.
The Licensee determined, through a combination of structural and systems analyses, that the consequences of a load drop will not preclude the ability to maintain ccre cooling. Licensee assumptions are'the same as those previously identified in Section 2.4.1 of this evaluation and are consistent, to a large degree, with the guidance of NUREG-0612.
It is noted that Appendix A of NUREG-0612 specifies that load analyses should consider a load drop from the maximum height at any point within the enrestrained movement of the crane.
In determining the limiting load drop for cach region identified within the Davis-Besse containment, however, the Licensee appears to have placed significant reliance on administrative controls which direct the movements of loads, such that certain load movements till not be conducted in certain a'reas of the containment.
In addition, credit appears to have been taken for other restrictions (i.e., lif t height) to reduce the effects of certain load drops. The Licensee appears to rely on the use of hdministrative controls to eliminate from furthur consideration certain heavy loads handled in the vicinity of 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 r::llability in reducing the likelihood of a load drop.
It is recognized, however, that in certain unique circumstances (specifically where the adminis-trative controls provide large separations between the control limits and the impact area of interest that are readily monitorable 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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TER-C5506-489 2.5.1.3 Conclusion Analyses performed by the Licensee partially demonstrate that core cooling can be maintained following a load drop in the containment onto equipment required for safe shutdown. However, additional,information is needed from the Licensee to identify and justify the administrative controls that are used to restrict movements of loads within the various regions, and 1
for which credit appears to have been taken in both the selection of the limiting load drop and to mitigate potential consequences of various load drops.
2.5.2 Component Cooling Pump Monorails 2.5.2.1 m==ary of Licensee Statements and Conclusions The limiting load drop for the component cooling water pump monorails is o drop of the component cooling water pump.
Based on results of the structural analysis, the Licensee concluded that "per'foration and scabbing were not probable" and any possible " effects were found to be insignificant." systems evaluation indicated that adequate physical separation exists so that suitable system redundancy would be retained and safe shutdown functions would not be lost.
2.5.2.2 Evaluation and Conclusion Analyses of a load drop by the component cooling water pump monorails indicated that existing design is adequate to satisfy Criterion IV of NUREG-0612, Section 5.1.
Assumptions used by the Licensee were generally consistent with those identified in NUREG-0612, Appendix A.
2.5.3 Intake Structure Gantry Crane 2.5.3.1 m==mry of Licensee Statements and Conclusions The limiting load drop in the service water intake structure area is a drop of the service water pump motors. Results of the structural analysis indicated that " perforation or scabbing were not probable," although it was ranklin Ibeerth Center
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TER-C5506-489 l
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considered in the systems evaluation. Systems analysis indicated that l
adequate physical separation exists between system components so that system redundancy would not be jeopardized and system functions would remain operable.
2.5.3.2 Evaluation and Conclusion Assumptions used by the Licensee to perform structural and systems analysis appear to be generally consistent with the intent of NUREG-0612.
Based upon analysis results, Criterion IV.of NUREG-0612, Section 5.,1 is,
satisfied.
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TER-C5506-489 3.
CDMCLUSIONS l
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 cf Section 2.
It is Provided to allow the reader to focus on the key topics that should be addressed in seeking to resolve issues where the degree of load handling reliability' provided by crane's at Davis-Besse Nuclear Power Station Unit I 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 NUEED 0612.
3.1 2NFONNhTION ISSUES i
i The information provided by the Licensee has been assessed as insufficient l
to support an independent conclusion that load handling reliability is consis-l tent with the evaluatice criteria of Section 2.1 in the following areas l
i Load Emndling System Evaluations (Sections 2.2.2 and 2.3.2.2)
The Licensee should evaluate the containment equipment jib cranes for i
compliance with the criteria of WUREG-0612.
In addition, an evaluation of the reactor service crane must be deferred since the crane is not in use and has not been evaluated by the Licensee.
i Spent Fuel Pool Interlocks (Section 2.3.1.3) i The Licensee should verify that adequate physical separation exists i
hetween the limits of crane travel and the spent fuel pool wall so that a dropped load will not impact, tip, or roll and cause damage to the spent
. fuel pool wall or spent fuel in the spent fuel pool.
Hovensets Nithin the Spent Fuel pool (Section 2.3.1.3)
Th allow movements of the pool divider gates, the Licensee should identify the means of enforcing plant procedures on a continuing basis to ensure that baron concentration in the spent fuel does not decrease to less than 1800 pse, as.well as identify load handling limitations if this limit is
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TER-C5506-489 violated.
In addition, the Licensee should provide necessary assurances that maximum enrichment will not exceed 3.5 weight percent.
Insd Drops in the Vicinity of the Beactor Vessel (Section 2.4.1.3)
To fully satisfy Criterion I, the Licensee should verify that a drop of the plenum assembly presents the bounding load drop for analyzing damage to the fuel and resultant offsite consequences.
3.2 APPROACE ISSUES This review has revealed the following issues wherein the approach or position taken by the Licensee, based on information provided thus far, is inconsistent with the staff's objectives as expressed in the evaluation criteria of Section 2.1.
Reactor Vessel Area Criticality Analysis (Section 2.4.1.3)
To demonstrate coupliance with Criterion II, the License should address the potential for exceeding limits for criticality specified in this criterion. Demonstration that a drop of a missile shield is not probable due to procedural controls is not adequate justification that Criterion II will not be exceeded.
R: actor Vessel Area Structural Integrity (Section 2.4.1.3)
The Licensee should reanalyze the consequences of a load drop onto the reactor vessel and its effects on structural integrity, using assumptions consistent with NUREG-0612, Appendix A, as well as reevaluate the consequences of new structural analyses on the systems analysis.
Imed Drops onto Components Required for Safe Shutdown (Section 2.5.1.3)
The Licensee appears to rely on the use of administrative controls to eliminate from further consideration certain heavy loads which are handled in various regions of the containment.
In general, such procedural con-trols 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.
It is recog-nised, however, that in certain unique cricumstances (specifically where the administrative controls provide large separations between the control limits and the impact area of interest which are readily monitorable 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. 1 am e m
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TER-C5506-489 3.
REFERENCES 1.
V. Stallo (NRC)
Letter to All Licensees
Subject:
Request for Additional Information on Control of Heavy Loads Near Spent Fuel May 17, 1978 2.
MRC NUREG-9612, " Control of Heavy Loads at Nuclear Power Plants",
July 1980 3.
D. G. Eisenhut (NRC)
Letter to All Operating Reactors
Subject:
Control of Heavy Imads December 22, 1980
- 4. 'FRC Technical Evaluation Report, " Control of Heavy Loads at Davis-Besse Nuclear power Station Unit 1" TER-C5506-348, August 9, 1983 5.
E. P. Crouse (TEC)
Letter to J. F. Stolz (NBC)
Subject:
Control of Heavy Ioads (Phase II)
June 10, 1983 6.
MRC NUREG-0544, " Single-Failure-Proof Cranes at Nuclear Power Plants" May 1979 7.
E. P. Crouse (TEC)
Subject Control of Heavy Icads (Phase I)
February 1, 1982 l
8.
FRC Draft Technical Evaluation Report, " Control of Heavy Loads at Davia-Desse Nuclear Power Station Unit 1" TER-C5506-530, August 27, 1983 9.
Code of Federal Regulations Energy (10CFR100) 10 - (Parts 0 tu 199)
January 1,1983 10.
Updated FSAR Davis-Besse Nuclear Power Station Unit 1 (Volume 12) n.a.n a r -