ML19344D868
| ML19344D868 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 07/31/1980 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | Peoples D COMMONWEALTH EDISON CO. |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, TASK-A-36, TASK-OR NUDOCS 8008260188 | |
| Download: ML19344D868 (25) | |
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Lw UNITED STATES 8 f NUCLEAR REGULATORY COMMISSION
- / g s.,(7 j.h WASHING TON, D. C. 20555
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July 31,1980 i
Docket Nos. 50-295 and 50-304 Mr. D. Louis Peoples Director of Nuclear Licensing Commonwealth Edison Company Post Office Box 767 Chicago, Illinois 60690
Dear Mr. Peoples:
In January 1978, the NRC published NUREG-0410 entitled, "NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants -
Report to Congress". As part of this program, the Task Action Plan for Unresolved Safety Issue Task No. A-36, " Control of Heavy Loads Near Spent Fuel," was issued.
We have completed our review of load handling operations at nuclear power plants. A report describing the results of this review will be issued in the near future as NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants - Resolution of TAP A-36."
This report contains several recomendations to be implemented by all licensees to assure the safe handling of heavy loads.
At the Indian Point Units 2 and 3, Zion Units 1 and 2, and Three Mile Island Unit 1 facilities, we are requesting licensee action to begin to implement these recomendations at this time on the schedule indicated in this letter.
To expedite your compliance with this request, we have enclosed the following:
1.
Guidelines for Control of Heavy Loads (Enclosure 1).
2.
Staff Position - Interim Actions for Control of Heavy Loads (Enclosure 2),
3.
Request for Additional Informatien on Control of Heavy Loads (Enclosure 3).
You are requested to review your controls for the handling of heavy loads to determine the extent to which the guidelines of Enclosure 1 are presently satisfied at your facility, and to identify the required changes and modifications in order to fully satisfy these p2idelines.
You are requested to implement the interim actions described in as soon as possible but no later than 90 days from the date of this letter.
THIS DOCUMENT CONTAINS P00R QUALITY PAGES 800826 0 / 5 g f
r Mr. D. Louis Peoples July 31,1980 You are further requested to submit a report documenting the results of your review and the required changes and modifications.
This report should include the information identified in Sections 2.1 through 2.4 of Enclosure 3, on how the guidelines of NUREG-0612 will be satisfied. This report should be submitted not later than the following schedule:
o Submit the Section 2.1 'information within three months from the date of this letter.
o Subrdt the Sections 2.2, 2.3 and 2.4 information within six months.
You should commence implementation of required changes and modifications as soon as possible without waiting on staff review, with the objective of completing changes, beyond the above interim actions, withir. two years of submittal of Section 2.4 for the above report.
Please notify your assigned NRC Project Manager if you will not be able to maintain these schedules.
i ncerely,
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Darrel G. gsenhut, Director Division of Licensing
Enclosures:
As stated cc: w/ enclosures See ne.wt page
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Mr. D. Louis Peoples Commonwealth Edison Corpany July 31,1980 cc: Robert J. Vollen, Esquire 109 North Dearborn Street Chicago, Illinois 60602 Dr. Cecil Lue-Hing Director of Research and Development Metropolitan Sanitary District of Greater Chicago 100 East Erie Street Chicago, Illinois 60611 Zion-Benton Public Library District 2600 Emmaus Avenue Zion, Illinois 60099 Mr. Phillip P. Steptoe Isham, Lincoln and Beale Counselors at Law One First National Plaza 42nd Floor Chicago, Illinois 60603 Susan N. Sekuler, Esquire Assistant Attorney General Environmental Control Division 188 West Randolph Street, Suite 2315 Chicago, Illinois 60601 U. S. Nuclear Regulatory Commission Resident Inspectors Office Post Office Box 288 Deerfield, Illinois 60015 O
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Encicsure (1) 5.
GUIDELINES FOR CONTROL OF HEAVY LOADS Our evaluation of the information provided by licensees indicates that existinc measures at operating plants to control the handling of heavy loads cover certain of the potential problem areas, but do not adequately cover the major causes of load handling accidents.
These major causes include operator errors, rigging failures, lack of adequate inspection and inadequate procedures.
The measures in effect vary from plant to plant, with some having detailed procedures while others do not, some have performed analyses of certain postulated load drops, certain plants have single-failure proof cranes, some PWR's have rapid containment isolation on high radiation, and many plants have technical speciff-cations that prohibit handling of heavy loads or a spent fuel cask over the scent fuel pool.
To provide adequata measures that minimize the occurrence of the principal causes of load nandling accicents and to provide an adequate level of defense-in-depth for handling of heavy loads near spent fuel and safe shutdown systems, the measures in effect should be upgraded.
5.1 Recommended Guidelines The following sections describe various alternative approaches which provide acceptable measures for the control of heavy loads.
The objectives of these.
guidelines are to assure that either (1) the potential for a loac crop is extremely small, or (2) for each area addressed, the following evaluation criteria are satisfied:
I.
Releases of radioactive material that may result from damage to spent fuel based on calculations involving accidental dropping of a postulated heavy load produce doses that are well within 10 CFR Part 100 limits of 300 rem thyroid, 25 rem whole body (analyses should show that doses are e;ual to or less than 1/4 of Part 100 limits);
II.
Damage to fuel and fuel storage racks based on calculations involving accidental dropping of a postulated heavy loac does not result in a configuration of the fuel such that k is larger than 0.05; gf III. Damage to the reactor vessel or the spent fuel pool based on calculations of damage following accicental cropping of a postulated neavy loac is limitec so as not to result in water leakage that could uncover the fuel, (makeup water provided to overcome laakage should be from a berated source of acequate concentration if the water being lost is borated); and IV.
Damage to ecuipment in recundant or dual safe snutccwn paths, based on calculations assuming tne accicental cro::cing of a ::cstulated neavy load, will be limited so as not to result in loss of required safe snutdown functions.
After reviewing tne nistorical cata available on crane ocerations, identifying the princical causes of loac crocs, anc considering ne type and frecuency of loac handling coerations at nuclear power clants, the NRC staff has cevelocec an overall philosochy tnat provides a defense-in-ce;:tn 3::proach for controlling the handling of heavy loacs.
This pnilosopny encompasses an intent to prevent as well as mitigate the consecuences of ::cstulatec accicental load crops.
The following summari es this cefense-in..ce::th ac::rcacn-0
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(1) Provide sufficient operator training, handling system design, load handling instructions, and equipment inspection to assure reliable operation of i
the handling system; and (2) Define safe load travel paths through procedures and operator training so that to the extent practical heavy loads avoid being carried over or near irradiated fuel or safe shutdown equipment; and (3) Provide mechanical stops or electrical interlocks to prevent movement of heavy loads over irradiated fuel or in proximity to equipment associated with redundant shutdown paths.
Certain alternative measures may be taken to comoensate for deficiencies in (2) and (3) above, such as the inability to prevent a particular heavy load from being brought over spent fuel (e.g., reactor vessel head).
These alterna-tive measures can include:
increasing crane reliability by providing dual load paths for certain components,_ increased-safety factors, and increased inspection as discussed in Section 5.1.6 of this report; restricting crane operations in the spent fuel pool area (PWRs) until fuel has decayed so that off-site releases would be sufficiently low if fuel were damaged; or analyzing the effects of postulated load drops to show that consequences are within acceptable limits.
Even if one of these alternative measures is selected, (1) and (2) acove should still be satisfied to provice maximum practical defense-in-depth.
The following sections provide guidelines on hcw the above defense-in-cepth approach may te satisfied for various plant areas.
Fault trees and associated probacilities were developed and used as cescribed in Bases for Guidelines, Section 5.2 of :nis report, to evaluate the adequacy of these guidelines and to assure a consistent level of protection for the various areas.
5.1.1 General All :lants nave overhead handling systems that are used to handle heavy loacs in the area of tne reacter vessel or spent fuel in the spent fuel pool.
Additionally, leads may be handled in other areas where their accidental drop ray damage safe shutdown systems.
Accordingly, all plants should satisfy each cf the following fer hancling heavy leads tnat coulc ce becugnt in proximity to or over safe shutcewn equipment or irradiated fuel in the scent fuel pool area and in containment (PWRs), in the reacter building (BWRs), and in other plant areas.
(1) Safe lead caths snculd be cefined for the =cvement of heavy 1cacs to minimi:e tne potential for heavy leacs, if crep;ed, to imcact irraciated fuel in the reacter vessel anc in the spent fuel pocl, or to impact safe snutccwn ecuiement.
The path shoulc follow, to the extent practical, structural ficer mem ers, beams, etc., suen tnat if the loac is dropped, the structure is more likely to withstand the im act.
These load paths should be cefinec in procecures, snewn en ecuipment layout drawings, and clearly marked on the floor in :ne area wnere the load is to be hanclec.
Oeviations from cefined lead caths should require written alternative procecures a;;revec by the plant safety review commi tee.
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(z) Procedures should be developed to cover load handling operations for heavy loads that are or could be handled over or in proximity to irradiated fuel or safe shutdown equipment.
At a minimum, procedures should cover handling of those loads listed in Table 3-1 of this report.
These procedures should include:
identification of required equipment; inspections and acceptance criteria required before movement of load; the steps and proper sequence to be followed in handling the load; defining the safe load path; and other special precautions.
(3) Crane ocerators should be, trained, qualified and conduct themselves in accordance with Chapter 2-3 of ANSI B30.2-1976, " Overhead and Gantry Cranes."
(4) Soecial liftino devices should satisfy the guidelines of ANSI N14.6-1978, "Stancara for Special Lif ting Devices for Shipping Containers Weighing 10,000 pounds (4500 kg) or More for Nuclear Materials." This standard should apply to all special lif ting devices which carry heavy loads in areas as defined above.
For operating plants certain inspections and load tests may be accepted in lieu of certain material requiremen'.s in the standard.
In addition, the stress design factor stated in Section 3.2.1.1 of ANSI N14.6 should be based on the combined maximum static and dynamic loads that could be imparted on the handling device based on characteristics of the crane which will be used." This is in lieu of the guideline in Section 3.2.1.1 of ANSI Ni4.6 which bases the stress design factor on only the weight (static load) of the load and of the intervening components of the special handling device.
(5) Lifting devices that are not scecially desianed should be installed and used in accordance witn the guicelines of ANSI S30.9-1971, " Slings."
However, in selecting the proper sling, the load used should be the sum of the static and maximum dynamic load.'
The rating identified on the sling should be in terms of the " static load" which produces the maximum static and dynamic load.
Where this restricts slings to use on only certain cranes, the slings should be clearly marked as to the cranes with whicn they may be used.
(6) The crane should be inspected, tested, and maintained in accordance with Cnacter 2-2 cf ANSI 530.2-1976, " Overhead ar.d Gantry Cranes," with the exception that tests and inspections should be performed prior to use where it is not practical to meet the frequencies of ANSI B30.2 for periodic inspection and test, or where frequency of crane use is less than the specified instection and test frequency (e.g., the polar crane insice a PWR containment may only be used every 12 to 18 months during refueling operations, and is generally not accessible during power operation.
ANSI S30.2, however, calls for certain inspections to be performed dsily or monthly.
For such cranes having limited usage, the inspections, tests, and maintenance should be performed prior to their use.)
For the purpose of selecting the proper sling, loads imoosec by the SSE need not ce included in the dynamic loacs imposec on the sling or lifting device.
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(7) The crane should be designed to meet the applicable criteria and guide-lines of Chapter 2-1 of ANSI B30.2-1976, " Overhead and Gantry Cranes" and of CMAA-70, " Specifications for Electric Overhead Travelling Cranes." An alternative to a specification in ANSI B30.2 or CHAA-70 may be accepted in lieu of specific compliance if the intent of the specification is satisfied.
5.1. 2 Spent Fuel Pool Area - PWR Many PWR's require that the spent fuel shipping cask be placed in the spent fuel pool for loading. Additionally, other heavy loads may be carried over or near the spent fuel pool using the overhead crane, including plant equipment, rad waste shipping casks, the damaged fuel container and replacement fuel storage racks.
Additionally, certain crane failures could cause the crane lower load block to be dropped, and there' 3re this should also be consioered as a heavy load.
The fuel handling crane is used for moving fuel and is generally not used for handling of heavy leads.
To provide assurance that the evaluation criteria of Section 5.1 are met for load handling operations in the spent fuel pool area, in addition to satisfying the general guidelines of Section 5.1.1, one of the following should be satisfied:
(1) The overhead crane and associated lifting dtvices used for handling heavy loads in the spent fuel pool area should satisfy the single-failure proof guidelines of Section 5.1.6 of this report.
OR (2) Each of the following is provided:
(a) Mechanical stops or electrical interlocks should be provided that prevent movement of the overhead crane load block over or within 15 feet horizontal (4.5 meters) of the spent fuel pool.
These mechanical stops or electrical interlocks should not be bypassed when the pool contains " hot" spent fuel, and should not be bypassed without approval from the shif t supervisor (or other designated plant management personnel).
The mechanical stops and electrical interlocks should be verified to be in place and operational prior to placing " hot" spent fuel in the pool.
(b) The mechanical stops or electrical interlocks'of 5.1.2(2)(a) above should also not be bypassed unless an analysis has demonstrated that damage cue to postulated load drops would not result in criticality or cause leakage that could uncover the fuel.
(c) To precluce rolling if drooped, the cask should not be carried at a height higher than necessary and in no case more than six (5) inches (15 cm) above the operating floor level of the refueling building or otner components and structures along the path of travel.
(c) Mechanical stops or electrical interlocks should be proviced to preclude crane travel from areas wnere a postulatec load droo could damage equipment from recuncant or alternate safe shutdown paths.
(e) Analyses shocic conform to the guidelines of Appencix A.
OR (3) Each of the following are provIded (Note:
This alternative is simlar to (a) above, exceo it allows movement of a heavy load, such as a cask, into the pool wnile it contains " hot" spent fuci if the pool is large enough to maintain wide separation between the load anc the " hot" spent fuel.):
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(a) " Hot" spent fuel should be concentrated f r. ut.e location in the spent fuel pool that is separated as much as possible from load paths.
(b) Mecha.-ical stops or electrical interlocks should be provided to prevent movement of the overhead crane load block over or within 25 feet horizontal (7.5 m) of the " hot" spent fuel.
To the extent practical, loads thould be moved over load paths that avoid the spent fuel pool and kept at least 25 feet (7.5 m) from the " hot" spent fuel unless necessary.
When it is necessary to bring loads within 25 feet of the restricted region, these mechanical stops or electrical interlocks should not be bypassed unless the spent fuel has decayed sufficiently as shown in Table 2.1-1 and 2.1-2, or unless the total inventory of gap activity for fuel within the protected area would result in offsite doses less than h of 10 CFR Part 100 if released, and such bypassing should require the approval from the shift supervisor (or other designated plant management individual).
The mechanical stops or electrical interlocks should be verified to be in place and operational prior to placing " hot" spent fuel in the pool.
(c) Mechanical stops or electrical interlocks should be provided to restrict crane travel from areas where a postulated load drop could camage equipment from recundant or alternate safe shutdown paths.
Analyses have demonstrated that a postulated load drop in any location not restricted by electrical interlocks or mechanical stops would not cause damage shat could result in criticality, cause leakage that could uncover the fuel, or cause loss of safe shutdown
. equipment.
(d) To preclude rolling, if dropped, the cask shoulc not be carried at a height higher than necessary and in no case more than six (6) inches (15 cm) above the operating floor level of the refueling building or other components and structures along the path of travel.
(e) Analyses snould conform to the guidelines of Accencix A.
OR (A) The effects of drops of heavy Icads should be analyzed and shown to satisfy the evaluation criteria of Section 5.1 of this report.
These analyses shoulc conform to the guidelines of Appendix A.
5.1.3 Containment suficing - pWR PWR containment buildings contain a polar crane that is used for removing and reinstalling shiele plugs, the reactor vessel neac, upper vessel internals, anc on occasion, otner neavy equipment such as the reactor coolant pumo, the reactor vessel inscection platform, and the cask used for damaged fuel.
Accitionally the crane load block may oe movec over fuel in the reactor wnen hancling smaller loacs or no load at all.
Due to the weign of the load block alone, this should also be consicered as a heavy load.
To provide assurance that the criteria of Section 5.1 are met for load handiing ocerations in the containment building, in accition to satisfying tne general guidelines of Section 5.1.1, one of the following should be satisfiec:
(1) The crane and asscciated lifting devices usec for handling heavy loacs in the containment buticing shoule satisfy the single-failure proof guicelines of Section 5.1.6 of :nis report.
05..
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(2) Rapid containment isolation is provided with prompt automatic actuation on high radiation so that postulated releases are within limits of evaluation Criterion I of Section 5.1 taking into account delay times in detection and actuation; and analyses have been performed to show that evaluation criteria II, III, and IV of Section 5.1 are satisfied for postulated load drops in this area.
These analyses should conform to the guidelines of Appendix A.
OR (3) The effects of drops of heavy Ioads should be analy:ed and shown to satisfy the evaluation criteria of Section 5.1.
Loads analy:ed should include the following:
reactor vessel head; upper vessel internals; vessel inspection platform; cask for damaged fuel; irradiated sample cask; reactor coolant pump; crane load block; and any other heavy loads brought over or near the reactor vessel or other equipment required for continued decay heat removal and maintaining shutdown.
In this analysis, credit may be taken for containment isolation if such is provided; however analyses should establish adequate detection and isolation time.
Addi-tionally, the analysis should conform to the guidelines of Appendix A.
5.1.4 Reactor Building - BWR The reactor building in BWRs typically contains the reactor vessel and spent fuel pool, as well as various safety-related equipment.
The reactor building overhead crane may be used in many day-to-day operations such as moving various shielded shipping casks or handling plant equipment related to maintenance or modification activities.
The crane is also used during refueling operations for removal and reinstallation of shield plugs, crywell head, reactor vessel head, steam dryers and separators, and refueling canal plugs and gates.
The crane would also be used subsequent to refueling for handling of the spent fuel shipping cask.
This cask may be lifted as high as 100 feet (30 m) above the grade elevation at whicn the cask is brougnt into the reactor building.
Additionally the overhead crane's load block may be moved over fuel in the reactor or over the spent fuel pool when handling smaller loads or no load at all.
Due to the weight of the load block alone, this should also be considered as a heavy load.
To assure that the evaluation criteria of Section 5.1 are satisfiec one of the following should be met in acdition to satisfying the general guicelines of Section 5.1.1:
(1) The reactor building crane, anc associated lifting cevices used for handling the acove neavy loads, shoulc satisfy the single-failure proof guicelines of Section 5.1.6 of this report, OR (2) The effects of heavy lead crois in the reactor building should be analy:ed to snow that the evaluation criteria of Section 5.1 are satisfied.
The loads analy:ed should include:
shield plugs, drywell head, reactor vessel head; steam dryers and separators; refueling canal plugs and gates; shielded spent fuel shi; ping casks; vessel inspection platform; anc any other heavy loads tnat may be Orcught over or near safe shutdown eouipment as well as fuel in :ne reactor vessel or the scent fuel pool.
Crecit may be taken in this ana,1ysis for coeration of tne Stancby Gas 5-6 i
Treatment System if facility technical specifications require its operation during periods when the load being analyzed would be handled.
The analysis should also conform to the guidelines of Appendix A.
5.1.5 Other Areas In other plant areas, loads may be handled which, if dropped in a certain Iccation, may damage safe shutdown equipment.
Although this is not a concern at all plants, loads that may damage safe shutdown equipment at some plants ir)clude the spent fuel shipping cask, turbine genarator parts in the turbine building, and plant equipment'such as pumps, motors, valves, heat exchangers, and switchgear.
Some of these leads may be less than the weight of a fuel assembly with its handling tool, but may be suf ficient to damage safe shutdown equipment.
(1)
If safe shutdown equipment are beneath or directly adjacent to a potential travel load path of overneac hanc' ling systems, (i.e., a path not restricted by limits of crane travel or by mechanical stops or electrical interlocks) one of the following should te satisfied in acdition to satisfying the general guidelines of Section 5.1.1:
(a) The crane and associated lifting devices should conform to the single-failure proof guicelines of Section 5.1.6 of this report; OR (b) If the load drop could 15 air the operation of equipment or cabling associated witn redunoant or dual safe shutdown paths, mechanical stops or electrical interlocks should be proviced to prevent movement of loads in proximity to these redundant or dual safe shutdown equipent (In this case credit should net be taken for intervening floors unless justifi.ed by analysis).
OR (c) The effects of load dropThave been analy:ed and the results indicate that damage to safe shutcewn equipment wculd not preclude operation of sufficient equicment to achieve safe shutcown.
Analyses should conform to the guidelines of Appendix A, as applicable.
(2) Where the safe shutdown equipment has a ceiling separating it from an overnead handling system, an alternative to Section 5.1.5(1) ateve woulc be to show by analysis that the largest postulatec icac handled oy the handling system would not penetrate the ceiling er cause spalling that could cause failure of the safe shutcown equipment.
5.1.6 Single-Failure-Proof Handling Systems i
For certain areas, to meet the guicelines of Sections 5.1.2, 5.1.3, 5.1.4, or 5.1.5, the alternative of upgracing the crane and lifting devices may ce chosen.
The purpose of the upgracing is to imoreve the reliability of the hancling system tnrough increased factors of safety and througn recundancy or duality in certain active components.
NUREG-0554, " Single-?ailure-Proof l
Cranes for Nuclear Power Plants," provices guidance for design, fa:rication, i
installation, and testing of new cranes tnat are of a hign reliability cesign.
For operating plants, Accendix C to this reocrt, "Mccification of Existing Cranes," :-revices guicelines on imolemer.tation of NUREG-0554 for coerating piants and plants uncer construction.'
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Section 5.1.1 of this report p*ovides certain guidance on slings and special handling devices. Where the alternative is chosen of upgrading the handling system to be " single-failure proof", then steps beyond the general guidelines of Section 5.1.1 should be taken.
Therefore, the following additional guidelines should be met where the alterna-tive of upgrading handling system reliability is chosen:
(1) Lifting Devices:
(a) Soecial lifting devices that are used for heavy loads in the area where the crane is to be upgraded should meet ANSI N14.6 1978,
" Standard For Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More For Nuclear Materials," as specified in Section 5.1.1(4) of this report except that the handling device should also comply with Section 6 of ANSI N14.6-1978.
If only a single lifting device is provided instead of dual devices, the special lifting device should have twice the design safety factor as required to satisfy the guidelines of Section 5.1.1(4).
- However, loads that have been evaluated and shown to satisfy the evaluation criteria of Section 5.1 need not have lif ting devices that also comply with Section 6 of ANSI N14.5.
(b) Liftinc devices that are not soecially desicned and that are used for hancling heavy loacs in the area wnere the crane is to be upgraded should meet ANSI B30.9 - 1971, " Slings" as specified in Section 5.1.l(5) of this rer rt, except that one of the following should also be satisfied unless the effects of a drop of the particular load have been analyzed and shown to satisfy the evaluation criteria of Section 5.1:
(i) Provide dual or recundant slings er lifting devices such that a single component failure or malfunction in the sling will not result in uncontrolled icwering of the load; OR (ii)InselectingtheproFersling,theloadusedshouldbetwice what is called for in meeting Section 5.1.1(5) of this repert.
(2) New cranes should te designed to meet NUREG-0554, " Single-Failure-Proof Cranes For Nuclear Pcwer Plants." For operating plants or plants under construction, the crane should be upgraded in accordance with the imple-mentation guidelines of Aopendix C of this recort.
(3) Interfacine lift ocints such as lifting lugs er cask trunions should also meet ene of tne foilewing for neavy leads handled in the area where tne crane is to be uograced unless the effects of a crep of the particular lead have been evaluated and shown to satisfy the evaluation criteria of Section 5.1:
(a) Provide redundancy er cuality such~ttat a single lift point failure will not result in uncentrollec lowering of the loac; lift peints should have a cesign safety facter with rescect to ultimate strength of five (5) times tne maximum comoinec concurrent static and dynamic loac after taking the singl,e lift point failure.
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(b) A non-redundant or non-dual lift point system should have a design safety factor of ten (10) times the maximum combined concurrent static and dynamic load.
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ENCLOSURE 2 STAFF POSITION -
INTERIM ACTIONS FOR CONTROL OF HEAVY LOADS
(~1) Safe load paths should be defined per the guidelines of Section 5.1.l(1) (See Enclosure'l);
(2)
Procedures should be developed and implemented per the guidelines of Section 5.1.1(2) (See Enclosure 1);
(3) Crane operators should be trained, qualified and conduct themselves per the guidelines of Section 5.1.1(3) (See Enclosure 1);
(a) Cranes should be inspected, tested, and maintained in accordance with the guidelines of Section 5.1.1(6) (See Enclosure 1); and (5)
In addition to the aoove, special attention should be given to procedures, equipment, and personnel for the handling of heavy loads over the core, such as vessel internals or vessel inspection tools. This special review should include the following for these loads:
(1) review of procedures for installation of rigging or lifting devices and movement of the load to assure that sufficient detail is provided and that instructions are clear and concise; (2) visual inspections of load bearing coerenents of cranes, slings, and special lifting devices to identify flaws or deficiencies that could lead to failure of the component; (3) appropriate repair and replacement of defective components; and (4) verify that the crane operators have been properly trained and are famiif ar with specific procedures used in handling these loads, e.g., hand signals, conduce of operations, and content of procedures.
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Enclosure (3)
REQUEST FOR ADDITIONAL INFORMATION ON CONTROL OF HEAVY LOADS 1.
INTRODUCTION Verification by the licensee that the risk associated with load-handling failures at nuclear power plants is extre=ely low will require a systa=atic evalua-tion of all load-handli:g systa=s at each site. The following specific information requests have been organi:ed to support such a systematic approach, and provide a basis for the staff's review of the licensee's evaluation. Additionally, they have
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been ergani:ed to address separately the two ha:ards requiring investigation (i.e.,
radiological censequences of damage to fuel and unavailability consequences of damage to certain systeas). The following gener.1 infor=ation is provided to assist in this evalua:1on and reduce the need for clarification as to :he intent and expect-ed results of this inquiry.
1.
Risk reduction can be de=enstrated by either of two approaches:
a.
The possibility of failure is extre=ely low due to handling-system design features (NCREG 0612, See:1on 5.1.6).
b.
The consequences of a failure can be shown te be acceptable (NUREG 0612, Section 5.1, Criteria I-IV).
Regardless of the approach selected, the general guidelines of NURIG 0612, Section 5.1.1, should be sa:isfied te provide maxi =u=
practical defense-in-depth.
2.
Evalua:Lons ccncerning radiological censequences er criticality safety, where used, can rely on ei:her the adoption of generic analyses reported in NUREG 0612, requiring only verfication that these generic assu=ptions are valid for a specific site, or e= ploy a si:e-specific analysis.
3.
Systens required for safe shu:down and centinued decay hea: removal are si:e-specific and are net, herefore, iden:ified in this request.
Individual plan:s shculd censider systems and components identified in Regula: cry Guide 1.29, Position C.1 (excep: :hese syste=s or por:1cas of systems tha: are required for (a) emergency core cooling, (b) pcs:-accident contain=en: hea: re=cval, or (c) post-acciden:
cen:ainmen: a: osphere cleanup), for evaluation and recogni:e tha:
the appreach :aken in :his respec: is similar to that identified in Regula:ory Guide 1.29, Posi:1ca C.2.
The fae: : hat a load-handling syste: :ay be prevented f rom opera:ing during plant condi:icas re-quiring the actual er potential use of se e of :hese systa=s, is re-
~
r cognized in this respect for infor ation.
4 The scope of this syste=atic review should include all heavy loads carried in areas where the potential for non-co=pliance with the acceptance criteria (NURIG 0612, Section 5.1) exists. A su==ary of typical loads to be considered has been provided in Attach =ent 6.
It is recog-ni:ed that so=e cranes vill carry additional =iscellaneous loads, some of which are not identifiable in detail in
- advance, la such cases an evaluation or analysis de=en-strating the acceptability of the handling of a range of loads should be provided.
5.
A some sites loads which =ust he evaluated vill include licensed shipping casks provided for :he transportatien of irradiated fuel, solidified radioac:1ve vas:e, spent resins, or other byproduct =aterial.
Licensing under 10CFR71 is not evidence tha: lifting devices for :hese shipping casks =eet the criteria specified in NURIG 0612, Jections 5.1.1 (4), 5.1.
1(5), 5.1.6(1), or 5.1.6(3), as appropriate, and thus does eli=inate the need to provide appropriate infor:a: ion not concerning these devices. A tabulation (A::ach=en: 7) is provided to indicate =ul:1ple-site use of these shipping casks.
The resul:s of the licensee's evaluatien, as reported in response to this request, should provide infernatien sufficien: for the s:sff to conduct an in-dependen: review to deter =ine cha: the inten: of this effore (i.e., the unif*or=
reduction of the potential hazard fro = load-handling-syste= failures) has been satisfied.
2.
INFORMATION REQUESTED FROM THE LICENSEE 2.1 GENERAL REQUIREMEWS FOR OVERHEAD HANDLING SYSTEMS NURIG 0612, See:1cn 3.1.1, iden:ifies several general guidelines rela:ed to
- he design and operatics of overhead load-handling systa=s in :he areas where spent fuel is stored,in :he vicinity of the reactor core, and in other areas of
- he plant where a load drop could result in da= age te equip =en: required for safe shu:dov: or decay heat re= oval.
Infor=ation provided in respense :o this see:1on should identify :he ex:ent of potentially ha:ardcus load-handling eperaticas at a si:e, :he ex:en: cf confer =ance :o appropria:a load-handling guidance, and :he changes required in c der :o confor= to the guidance.
l 1.
Reper: the results of ycur review cf plan: arrange =en:s to iden:1fy all everhead handling systens fr== vhich a load drop may resul: in da= age to any syste required for plan:
shu:deve er decay haa: re= eval (taking no credi: for any
_l.
i I
interlocks, technical specifications, operating procedures, or detailed structural analysis).
2.
Justify the exclusion of any overhead handling systa= fre=
the above category by verifying : hat there is sufficient physical separation frc= any lead-i= pact point and any safety-rela:ed co=ponent to pernit a de:er=ination by inspec-tion that no heavy load drop can result in da= age to any systa= or ce=ponent required for plant shutdown or core decay heat removal.
3.
With respect to the design and operation of heavy-load-handling syste=s in the containnen: and the spent-fuel-pool area and those Icad-handling syste=s iden:ified in 2.1-1, above, provide your evaluation concerning ce=pliance with the guidelines of NURIG 0612, See: ion 5.1.1.
The following specific infor=ation should be included in yeur reply:
a.
Drawings or sketches sufficient to clearly identify :he loca:icn of safe lead paths, spent fuel, and safety-related equip =ent.
b.
A discussion of measures taken :o ensure that lead-handling operations re=ain within safe lead paths, including procedures, if any, for deviation fec= these paths, c.
A *abulation of heavy loads to be handled by each cre.ne.which includes :he load identification, load weight, its designa:ed liftinF device, and verifi-ca: ion that the handling cf such lead is governed by a vri::en procedure cen:aining, as a =ini=u=,
the infer =ation identified in NURIG 0612, Section 5.1.1(2).
d.
Verifica: ion tha: lif:ing devices iden:ified in 2.1.
3-c, above, ce= ply vi:h the require =ents of ANSI 14 6-1973, er ANSI 330.9-1971 as appropria:e. yor lif:ing devices where these standards, as supple =ented by NURIG 0612, See:ica 5.1.1(4) or 5.1.1(5), are nor
=et, describe any propcsed alternatives and de=en-stra:e their equivalency in :er=s of lead-handling reliability.
e.
Verifica:ica : hat ANSI 330.2-1976, Chap:er 2-2, has been invoked with respec: :: crane inspectien, testing, and =aintenance. Where any excep:1cn is taken to this standard, suf ficie-- ' '----tien should be provided :o de= ens::a:e :he equivalency of proposed alternatives.
f.
Verifica:ica : hat crane design c:= plies vi:h the guide-lines of CMAA Specifica:icn 70 and Chap:ar 2-1 ef ANSI 330.2-1976, including :he de= ens:: :ica Of equivalency of actual design require =ents for instances where spe-cific ce=pliance vi:h :hese s:andards is no: provided.
4.
=
g.
Exceptions, if any, taken to ANSI 330.2-1976 vith respect to cperator training, qualifica: ion, and co nduct.
2.2 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN THE VICINITY OF FUEL STORAGE POOLS NURIG 0612, section 5.1.2, provides guidelines concerning :he design and opera:1cn of load-handling syste=s in :he vicini:7 of stored, spen: fuel.
Infor:ation provided in respcase to this section shculd de=ons: rate tha: ade-quate =easures have been :aken :o a:sure tha: in this area, either the likeli-hood of a load drop which =1ght da= age spent fuel is extre=ely r=all, or tha:
the esti=ated consequences of such a drop vill not exceed :he li=its set by the evalua:ica criteria of NURIG 0512, See:ica 5.1, Cri:eria I thrcugh III.
1 Identify by na:e, type, capacity, and equip =ent _esignator, any cranes physically capable (i.e., ignoring interlocks,
=oveable =echanical s: ops, or opera:ing procedures) of carry-ing loads which could, if dropped, la:d or f all in:o :he spect fuel pool.
2.
Justify :he exclusica of any cranes in this area fro = the above category by verifying :ha: : hey are incapable of carrying heavy loads or are per=cnently prevented frc= move-
=ent of the hook centerline closer than 15 feet to the pool boundary, er by providing a suita:1e analysis de=enstrating tha: for any failure =cde, no heasy load can fall into the fuel-s:crage pool.
3.
Iden:ify any cranes listed in 2.2-1, above, which you have evalua:ed as having suf ficie:: design features to =ake :he likeliheed of a load drop ex:re=ely r=all for all loads :o be carried and the basis for :his evaluatien (i.e., ce=plete ec=pliance vi h NURIG 0612, See:ict 5.1.6 er partial ccm-pliance supple =en:ed by sui:able alterna:1ve or additional design features). yer each crane se evalua:ed, provide the lead-handli:g-syste= (i.e., crane-1 cad-ce=bina:1on) infor:a-
- ica specified in A::ach=en: 1.
1 Fer cranes identified in 2.0-1, aaeve, ne: categoriced accord-ing :o 2.0-3, de=enstra:e : hat tre cri:eria of NUEIG 0612, See:icn 5.1, are satisfied. Cc=p'iance with Criterien IV will be de=onstra:ed in respense :c Section 2.4 ef :his request. 21:h respect :c Cri:eria ! :hrough III, provide a discussion of ycur evalua:1cn of cra:e operacien 1: :he spent fuel area and your de:ertinatics of co=pliance. This respense should include the fe'.lcuing infer:atien fer each crane:
a.
Which al:ernatives (e.g.,
2, 2. er - frc= :hese iden:ified in NURIG Oel, Sectica.. '_.2, have been selected.
-;r
i
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b.-
If Alternative 2 or 3 is selected, discuss the crane =otion li=itacien i= posed by electrical interlocks or mechanical stops and indicate the circu= stances, if any, under which these protective devices =ay be bypassed or re=oved.
Discuss any ad=inistrative procedures invoked to ensure proper au:horization of bypass or re= oval, and provide any related or proposed technical specification (operational and surveillance) provided to ensure the operability.of such electrical in:ericcks or
=echanical stops.
c.
Where reliance is placed on crane operational limi:a:1ons with respect to the ti=e of the storage of certain quantities of spent fuel at specific post-irradiation decay times, provide present and/or proposed technical specifications and discuss administrative or physical controls provided to ensure that these assu=ptions re=ain valid.
d.
Where reliance is placed on the physical location of specific fuel =odules at certain post-irradiation decay times, previde present and/or proposed techni-cal specifica: ions r.nd discuss administrative or physical controls provided to ensure tha: these assu=ptions re=ain valid, e.
Analyses perfor=ed to de=enstrate ec=pliance with Criteria I through III should confor= to the guide-lines of Attachnent 5.
Justify any exception taken to these guidelines, and provide the specific infor-cation requested in A::ach=en: 2, 3, or 4, as appro-priate, for each analysis perfor=ed.
2.3 SPECIFIC REQUIREMENTS OF 1RHEAD HANDLING SYSTEMS OPERATING IN THE CCNTAINMENT NURIG 0612, Section 5.1.3, provides guidelines concerning the design and opera: ion of load-handling systems in the vicinity of the reac:cr core.
Infor-
=ation provided in response :o this section should be sufficent to de=enstrate.
tha: adequate ceasures have been :aken to ensure :ha: in this area, either :he likeliheed of a lead drop which might damage spent fuel is extre=ely s=all, or
- ha: the es:i=a:ed consequences of such a drep vill not exceed :he li=its set by :he evaluation cri:eria of NURIG 0612, Section 5.1, Criteria I :hrough III.
1.
Identify by na=e, :ype, capacity, and equipment designator, any cranes physically capable (i.e.,
taking no credit for any interlocks or cpera:ing procedures) of carrying heavy leads ever :he reactor vessel.
-5 '
i
2.
Justify the exclusion of any cranes in this area fro = the above category by verifying that they are incapable of carrying heavy loads, or are permanently prevented fro:
the =ove=ent of any load either directly over the reactor vessel or to such a location where in the event of any load-handling-system failure, the load =ay land in or on the reactor vessel.
3.
Identify any cranes listed in 2.3-1, above, which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely s=all for all loads to be carried and the basis for this evaluation (i.e., co=-
plete compliance with NURIG 0612, Section 5.1.6, or partial ec=pliance supplemented by suitable al:ernative or additional design features). For each crane so evalua:ed, provide the lead-handling-systa= (i.e., crane-load-combination) informa-tion specified in Attachment 1.
4.
For cranes identified in 2.3-1, above, not categorized accord-ing to 2.3-3, de=enstrate that the evaluation criteria of NURIG 0612, Section 5.1, are sa:irfied. Co=pliance with Criterion IV will be demonsarated in your response to Sec-tien 2.4 of this request. With respec: to Criteria I through III, provide a discussion of your evaluation of crane opera-tion in the con:ain=ent and your deter =ination of co=pliance.
This response should include the following infor=a: ion for each crane:
a.
Where reliance is placed on the ins allation and use of electrical inc.erlocks or methanical stops, indicate the circu= stances under which these protec:ive devices can be re=oved or bypassed and :he ad=instrative pro-cedures invoked to ensure proper authori:ation of such action. Discuss any rela:ed or proposed technical specification concerning the bypassing of such interlocks.
b.
Where reliance is placed on c:her, site-specific con-sidera:icns (e.g., refueling sequencing), provide present or proposed technical specifications and dis-cuss ad=ints::a:ive or physical controls provided to ensure the continued validity of such considera: ices, c.
Analyses perferned to de= ens:: ate ::=pliance with Cri:eria I threugh III shculd confor= wi:h :he guide-lines of Attachnent 5.
Justify any exception taken to :hese guidelines, and provide the specific infor-
=a: ion recuested in A::ach=en: 2, 3, or 4, as appro-pria:e, for each analysis perfor=ed.
2.4 SPECIF:C REOUIREMENTS FOR OVERHEAD HANDLING SYSTE.MS OPERATING IN PLANT AREAS CCNTAINING EQUIPMENT REOUIRED FCR REACTOR SHUTDOWN, CCRE DECAY HEAT REMOVAL, OR SPENT FUEL POOL C0 CLING NCRIG C612, See: ion 3.1.5, provides guidelines cencerning the design and opera: ion of lead-handling syste=s in'the vicinity of equipment or ce=ponents t
-o-
1 i
i required for' safe. reac:or shutdown and decay heat removal.
Infor=ation pro-vided in response to this section should be sufficient to de=or. strate that adequate =easures have been taken to ensure that in these areas, either the likelihood of a load drop which might prevent safe reactor shutdown or prohibit continued decay heat re= oval is extre=ely s=all, or that da= age to such equip-
=et.: from load drops will be limited in order not to result in the loss of these safety-related functions. Cranes which must be evaluated in this section have been previously identified in your response to 2.1-1, and their loads in your response to 2.1-3-c.
1.
Identify any cranes listed in 2.1-1, above, which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely r=all for all loads to be carried and the basis for this evaluation (i.e., complete ce=pliance with NUREG 0612, Section 5.1.6, or partial com-liance supplemented by suitable alter:acive or additional design features). For each crane so evaluated, provide the lead-handling-syste= (i.e., crane-load-co=bination) tr. forma-tion specified in Attachment 1.
2.
For any cranes identified in 2.1-1 not designated as single-failure-proof in 2.4-1, a comprehensive ha:ard evaluation should be provided which includes the following infor=ation:
a.
The presen:ation in a = atrix for=at of all heavy loads and potential i= pact areas where da= age might occur to safety-related equipment.
Heavy loads identification should include designation and weight or cross-reference :o infor=ation pro-vided in 2.1-3-c.
I= pact areas should be identi-fied by construe: ion :enes and elevations or by some other method such tha: the i= pact area can be located on the plant general arrange =en: drawings.
Figure 1 provides a typical =a: ix.
b.
For each interaction identified, indicate which of the load and i= pact area combinations can be el1=inated because of separation and redundancy of safety-related equip =ent, =echanical stops and/or electrical in:erlocks, or other site-specific considerations.
Ilimination on the basis of the afore=entioned considera:1ons should be supple =ented by :he following specific infer:ation:
(1) For load /:arge: ce=binaticas eli=isa:ed because of separation and redundancy of safe:y-rela:ed equip =ent, discuss :he basis for de:er=ining that load drops will not affec: continued system operation (i.e.,
the ability of the sys:e= to perfor= its safety-related function).
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s
=
(2) Where sechanical steps or electrical inter-locks are te be provided, present details showing the areas where crane travel vill be prohibited. Additionally, provide a discus-sion concerning the procedures that are to be used for authoriting the bypassing of interlocks cr re=ovable stops, for verifying that interlocks are functional prior to crane use, and for verifying that interlocks are restored to operability after operations which require bypassing have been co=pleted.
(3) Where load / target combinations are eli=inated on the basis of other, site-specific consi-derations (e.g., =aintenance sequencing), pro-vide present and/or proposed technical speci-fications and discuss administrative procedures or physical constraints invoked to ensure the continued validity of such considerations.
c.
For interactions not eli=inated by the analysis of 2.4-2-b, above, identify a=y handling systems for specific loads which you have evaluated as having sufficient design fea-tures to make the likelihood of a load drop extremely small and the basis for this evaluation (i.e., complete compliance with NUREG 0612, Sect Wn 5.1.6, or partial co=pliance sup-plemented by suitable alternative or additional design fea-cures).
For each crana so evaluated, provide the load-handling-syste= (i.e., :rane-load-co=bination) infornation specified in Attachment 1.
d.
For interactions not eli=inated in 2.4-2-b or 2.4-2-c, above, de=enstrate using appropriate analysis that da= age vould not preclude operation of sufficient equip =ent to allow the syste= to perfor= its safety function following a load drop (NURIG 0612, Section 5.1, Criterion IV).
For each analysis so cenducted, the following information shculd be provided.
(1) An indicatien of whether or not, for the specific lead be.ng investigated, the over-d head crane-handling syste= is designed and constructed such that the hoisting syste=
vill retain its ?. cad in the event of seis=1c accelerations eq.ivalent to those of a safe shutdown earthquake (SSE).
(2) The basis for any exceptions taken to the analytical guidelines of Attach =ent 5.
(3) The infer =ati;n requested in Attach =ent 4.
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NOTES TO yIGURI 1 Note 1:
Indicate by sy=bols the safety-related equipment.
The licensee should provide a list consistent with the clarification provided in 1.2-3.
Note 2: Ha:ard Eli=ination Categories Crane travel for this area / load co=bination prohibited a.
by electrical interlocks or =echanical stops.
b.
Systa= redundancy and separation precludes lors of capability of syste= to perfor its safety-related function following this load drop in this area.
Site-specific considerations eli=inate the need to con-c.
sider load / equip =ent co=bination.
d.
Likelihood of handling systa= failure for this load is extremely small (i.e. section 5.1.6 NURIG 0612 satis-fied).
Analysis de=enstrates that crane failure and load drop e.
vill not da= age saf ety-related equip =ent.
O S
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Attachment (1)
SINGLE-FAILUF.E-PROOF HANDLING SYSTEMS 1.
Provide the na e of the =anufacturer and the design-rated lead (DRL).
If
- he =axi=u= critical load (MCL), as defined in NURIG 0554, is not the sa=e as the DRL, provide this capacity.
2.
Provide a de: ailed evalua:ica of the overhead handling systa= with respect to the features of design, fabrica:ics, inspection, testing, and operacion as delinea:ed in NURIG 0554 and supple =ented by :Ee iden:ified alternatives specified in NURIG 0612, Appendix C.
Uhis evalua:ic: =us: include a poin:-
by-pois: ce=parison for each section of NURIG 0534 If the alternatives of NURIG 0612, Appendix C, are used for certain applica:1ons in lieu of c:= plying vi:h :he rece==enda: ion of NURIG 0554, this should be explici:ly sta:ed. If an alternative to any of those c=n:ained in NURIG 0554 or NURIG C612, Appendix C, is prepcsed details =ust be previded on the proposed alterna:1ve to de=enstrate its equivalency.
3.
'a'ich respect :o the seis=ic analysis e= ployed to de=enstra:e :ha: :he over-head handling sys:e= can re:ain the load during a seis=ic event equal to a safe shutdown earthquake, provide a descrip: ion of the =ethod of analys,is, the assu=p: ions used, and the = ache =atical =edel evalca:ed in :he a=alysis.
The description of assu=pticas should include the basis for selection of
- rolley and load posi:icn.
4 Provide an evaluatics of :he lif:ing devices for each single-failure-proof handling syste= wi:h respec: to the guidelines of NURIG 0612, Sention 5.1.6.
5.
Provide an evaluatien of the interf acing lif t poin:s with respect :o :he guidelines of NURIG 0612, See:ica 5.1.6.
9
.