ML20041C541

Provides Response to NRC 801222 Request for Results of Heavy Load Review & Required Mods.Insp & Preventive Maint Program for Reactor Containment Bldg Polar Crane Will Minimize Potential for Load Drops

ML20041C541
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	02/24/1982
From:	Linder F DAIRYLAND POWER COOPERATIVE
To:	Crutchfield D Office of Nuclear Reactor Regulation
References
REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR LAC-8114, NUDOCS 8203020229
Download: ML20041C541 (20)

Text

{{#Wiki_filter:* ,e D DA/RYLAND k [k COOPERA T/VE e o. box Bi7 2615 EAST AV SOUTH LA CROSSE. WISCONSIN S4601 (608) 788-4000 February 24, 1982 In reply, please refer to LAC-8114 Director of Nuclear Reactor Regulation ATTN: Mr. Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 b Division of Operating Reactors qg g0000 U. S. Nuclear Regulatory Commission ~ Washington, D. C. 20555 g gg g SUDJECT: DAIRYLAND POWER COOPERATIVE h LA CROSSE BOILING WATER REACTOR (LACBWR) nu PROVISIONAL OPERATING LICENSE NO. DPR-45 CONTROL OF HEAVY LOADS N 4

Reference:

(1) NRC Letter, Eisenhut to All Licensees of Operating Plants and Applicants for Oper-ating Licenses and Holders of Construction Permits, dated December 22, 1980. (2) DPC Letter, LAC-7573, Linder to Crutchfield, dated June 1, 1981. (3) DPC Letter, LAC-8031, Linder to Crutchfield, dated January 19, 1982. Gentlemen: Reference (2) addressed the interim actions of Enclosure 2 of Reference (1). Reference (1) requested further licensee actions in the form of a report documenting the results of review of subject and required changes and modifications. The following information is hereby submitted in response to this request. The item numbers are those sections from Enclosure 3 of Reference (1). 2.1-1 Report the results of your revicu of plant arrangements to identify att overhead handling systems from which a load drop may result in damage to any system required for plant shutdoun or decay heat removal (taking credit for any interlocks, technical specifications, operating procelures, or detailed structural analysis). DPC RESPONSE: A review of the plant arrangement shows that there is an overhead handling system, a 50-ton capacity polar crane, in the reactor containment building which may have a potential for a load drop which could result in damage to the reactor decay heat system and/or control drive mechanisms used for reactor shutdown. ( $8 ) 8203020229 820224 I PDR ADOCK 05000409 P PDR

Mr. Dennis M. Crutchfield, Chief LAC-8114 Operating Reactors Branch No. 5 February 24, 1982 Interlocks, technical specifications, operating procedures and detailed structural analysis all exist which reduces this potential. No other plant cranes fall within the purview of Reference (1). 2.1-2 Justify the exclusion of any overhead handling system from the above category by verifying that there is sufficient phyoical ocparation from any load-impact point and any safety-related component to permit a determination by inapection that no heavy load drop can result in damage to any cystem or component re-quired for plant chutdoun or decay heat removal. DPC RESPONSE: The fuel transfer bridge contains a hoisting mechanism for handling fuel elements,both new and irradiated, and control rodt, The hoisting mechanism is physically limited to that of a fuel element or less. There is a bridge crane located overhead in the turbine building, and a gantry crane located outside. Neither the turbine building nor outside crane are capable of a load drop in the proximity of any system or component required for plant shutdown or decay heat removal. 2.1-3. Wit l: respect to the design and operation of heavy-load-handling systema in the reactor building and those load-handling systems identified in 2.1-1, above, provide your evaluation concerning compliance with the guidelines of NUREG-0612, Section 5.1.1. The follouing specific infor-mation should be included in your reply: Drauings or sketches sufficient to clearly a. identify the locationof cafe load paths, opent fuel, and safety-related equipment. DPC RESPONSE: Enclosed are several drawings, Figures 6.1 through 6.6, which identify the safe load path of the fuel shipping cask, location of spent fuel pool, reactor vessel head and shield plug set down area, canal plug and its storage area when refueling, also the decay heat pump and heat exchanger. The piping for the decay heat enters and exits the Forced Circulating Pump (FCP) cubicle where it connects to the 1A FCP loop piping..

Mr. Dennis M. Crutchfield, Chief LAC-8114 Operating Reactors Branch No. 5 February 24, 1982 2.1-3.b. A discussion of measures taken to ensure that load-handling operations remain within safe load paths, including procedures, if any, for deviation from these paths. DPC RESPONSE: TABLE 1. REACTOR BUILDING POLAR CRANE DESIGNATED LIFTING PROCEDURE LOAD WEIGHT DEVICE 1. Reactor Vessel 12.5 T YES Vol. VI, Sect. 4 Head 2. Insulation 1T YES Vol. VI, Sect. 3 Cover 3. Head Piping 1/2 T YES Vol. VI, Sect. 3 4. Shield Plugs 30T,31T, YES Vol. VI, Sect. 2 (3 Total) 32T 5. Transfer Canal 36 T YES Vol. VI, Sect. 6 Shield Plug 6. Transfer Canal 1/4 T YES Vol. VI, Sect. 7 Gate 7. Core Spray 1T YES Vol. VI, Sect. 5 Bundle 8. Fuel Shipping 35 T YES (2) Special Procedure Cask 9. Other Casks 2-7 T YES (2) Special Procedure 10. Crane Load 2T NOT Not Applicable Block APPLICABLE 11. FCP 14 T YES (2) Special Procedure i (1) See 2.1-2 for exclusion of other cranes. (2) Lifting devices are designated in special procedure. Mr. Dennis M. Crutchfield, Chief LAC-8114 Operating Reactors Branch No. 5 February 24, 1982 Load handling operations are conducted under reviewed and approved procedures, on which operators are trained or briefed. All operations involving the reactor vessel head,. shield plugs, insulation cover, head piping, canal plug, are contained in procedures in the LACBWR Operating Manual, Volume VI. Special procedures were developed for use with the irradiated fuel shipping cask, and other casks that have been handled in the reactor containment building. Also, special procedures were developed for lifting the FCP to the 701' floor in containment, designating specific load travel paths, set-down locations, lifting rigs required, personnel designated to supervise and perform various tasks in the handling operation, inspections and tests required, and plant condi-tions during the lift. 2.1-3.c. A tabulation of heavy loado to be handled by each crane uhich includes the toad identification, load ucight, its designated lifting device, and verifi-cation that the handling of such load is governed by a uritten procedure containing, ao a minimum, the information identifed in NUREG-0612, Section 5.1.1(2). DPC RESPONSE: The procedures listed in Table 1, each contain: identification of required equipment, inspectior s and acceptance of equipment as pre-requisites, the sequence of the lift, load path to be followed, pre-cautions and-other instructions, all approved in accordance with administrative procedures. 2.1-3.d. Verifica tion that lifting deviceo identified in 2.1. 3-c, above, comply uith the requirements of ANSI N14. 6-1978, or ANSI B30.9-1971 as appropriate. For lift-ing deviceo ahere these standarda, as supplemented by NUREG-OG12, Section 5.1.1(4) or 5.1.1 (5), are not met, describe any proposed atternatives and demon-strate their equivalency in terms of toad-handling reliability. DPC RESPONSE: Prior to use on the site, lifting devices for casks are verified to be able to perform the lift desired by requiring certain tests, inspections, material certifications and design data, as appropriate, from the manufacturer. Additionally, a load test is performed on Mr. Dennis M. Crutchfield, Chief LAC-8114 Operating Reactors Branch No. 5 February 24, 1982 the lifting device on site. These actions assure conformance to the guidelines of ANSI N14.6-1978. Lifting devices used for other loads on Table 1, except Insulation Cover and FCP, are of a unique design and are used only on that load. These lifting devices are inspected prior to load lift, and immediately upon applying the load for each lift. Lifti.ng devices for the Insulation Cover and FCP are commercial wire rope slings, and conform completely with ANSI B30.9-1971. Our inspection and testing program for slings includes these lifting slings. 2.1-3.c Verification that ANSI B30.2-1976, Chapter 2-2, has been invoked with respect to crane inspec-tion, testing, and maintenance. Where any excep-tion is taken to this standard, sufficient informa-tion should be provided to demonstrate the equiv-atency of proposed alternatives. DPC RESPONSE: All of the elements in Chapter 2-2 of ANSI B30.2-1976 with respect to inspection, testing and maintenance, are incorporated in the preventive maintenance procedure M-37 which is in use at LACBWR. 2.1-3. f Verification that crane design complies with the guide-lines of CMAA Specification 70 and Chapter 2-1 of ANSI B30.2-1976, including the demonstration of equivalency l of actual design requirements for instances where specific compliance uith these standards is not pro-vided. t l DPC RESPONSE: l l The reactor containment building crane was designed and constructed l to Allis-Chalmers procurement Specification 41-552 (Sargent and l Lundy Specification W-1759). The specifications therein have been compared to those in Chapter 2-1 of ANSI B30.2-1976 and found to be j equivalent. i ;

Mr. Dennis M. Crutchfield, Chief LAC-8114 Operating Reactors Branch No. 5 February 24, 1982 2.1-3.g Exceptions, if any, taken to ANSI B30.2-1976 uith respect to operator training, qualification, and conduct. DPC RESPONSE: LACBWR crane operators are qualified to LACBWR Administrative Control Procedure 23.1, "LACBWR Crane Operator Qualification and Certifica-tion", which encompasses the requirements of ANSI B30.2-1976 with respect to operator training, qualification and conduct.

2. 2 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN THE REACTOR BUILDING NUREG 0612, Section 5.1.4, provides guidelines concerning the design and operation of load-handling systems in the vicinity of spent fuel in the reactor vessel or in storage.

Information pro-vided in response to this section should demonstrate that adequate measures have been taken to ensure that, in this area, either the likelihood of a load drop which might damage spent fuel is extremely small, or that the estimated consequences of such a drop vill not exceed the limits set by the evaluation criteria of NUREG 0612, Section 5.1, Criteria I through III. 1. Identify by name, type, capacity, and equipment designator, any cranes physically _ capable (i.e., ignoring interlocks, moveable mechanical stops, or operating procedures) of carrying loads over spent fuel in the storage pool or in the reactor vesscZ. DPC RESPONSE: NAME: Reactor Containment Building TYPE: Polar Bridge CAPACITY: 50-Ton Main Hoist, 5-Ton Auxiliary Hoist EQUIPMENT DESIGNATOR: This is the same crane described in response to 2.1-1. Mr. Dennis M. Crutchfield, Chief LAC-8114 Operating Reactors Branch No. 5 February 24, 1982 2. Justify the exclusion of any cranes in this area from the above category by verifying that they are incapable of carrying heavy loads or are permanently prevented from movement of heavy loads over stored fuel or into any location uhere, follouing any failure, such load may drop into the reactor vessel or spent fuel storage pool DPC RESPONSE: No cranes other than the Reactor Containment Building crane, are in this category. 3. Identify any cranes listed in 2.2-1, above, which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely small for all loads to be carried and the basis for this eval-uation (i.e., complete compliance uith NUREG-0612, Section 5.1.6 or partial compliance supplemented by suitable altern2tive or additional design features). For each crane 30 evaluated, provide the load-handling-system (i.e., crane-load-combination) information specified in Attachment 1. DPC RESPONSE: The crane identified in 2.2-1, the Reactor Containment Building crane, is not-designed nor upgraded to be a single-failure-proof load handling

system, i.e.,

is not in complete compliance with NUREG-0612, Section 5.1.6. 4. For cranes identified in 2.2-1, above, not categorized according to 2.2-3, demonstrate that the criteria of NUREG-0612, Section 5.1, are satisfied. Compliance with Criterion IV uilt be demonstrated in response to Section 2.4 of this request. With respect to Criteria I through III, provide a discussion of your evaluation of crane operation in the Reactor Building and your determination of compliance. This response should include the following information for each crane:.

l. 1 i Mr. Dennis M. Crutchfield, Chief LAC-8114 Operating Reactors Branch No. 5 February 24,.1982 Where reliance is placed on the instattation a. and use of electrical interlocks or mechanical stops, indicate the circumstances under which thece protective devices can be removed or by-i passed and the administrative procedures invoked to ensure proper authorizationof such action. Discuss any related or proposed technical spec-ificationo concerning the bypass of such interlocks. DPC RESPONSE: i In response to the issue of Criterion I, Amendment No. 18 to Pro-visional Operating License No. DPR-45 for the La Crosse Boiling Water Reactor, dated February 4, 1980, contained Section 3.6.2, ) which referred to the Safety Evaluation of the potential consequences j of a drop of the spent fuel shipping cask into the spent fuel element storage well, issued by the Nuclear Regulatory Commission dated October 22, 1975 in conjunction with Amendment No. 4 to POR No. DPR-45 dated March 18, 1975. These two amendments concluded that for the consequences of a drop of a 50-ton (maximum capability of the crane) cask into the fuel element storage well pool with a freshly discharged core in the pool to be well within 10 CFR Part 100 i exposure guidelines, the' containment must be isolated if the spent fuel has decayed less than 43 days. On the rare occasion when.the entire core may be off loaded from the reactor vessel into the FESW, the required decay time will be extended to 51 days to compensate for the reduced depth of water (16 feet) above the stored spent fuel i elements. The Office of Nuclear Reactor Regulation specified administrative procedures in the Safety Evaluation of October 22, 1975 which require that the cask not be lifted more than six inches from the operating i floor (701 foot), and control the path of cask movement by both bridge l and trolley interlocks. These interlocks are effected prior to the movement of the cask over the concentric route between storage pool and hatch area, see Figure 6.6. The procedure developed for the spent. fuel cask handling specifies when.these protective actions are j. l imposed and removed. I Technical Specification 4.2.1.9 imposed requirements for containment isolation when handling a shipping cask (heavy load) and also for-pool water level. No new technical specifications or changes are contemplated or proposed. ]

I Mr. Dennis M. Crutchfield, Chief LAC-8114 Operating Reactors Branch No. 5 February 24, 1982 s In response to the issue of Criterion II, Amendment No. 18 to Provisional Operating License No. DPR-45, dated February 4,

1980, contained Section 3.1 which concludes that when any number of. fuel assemblies, with no more than 22.6 grams of Uranium-235 per axial centimeter of fuel assembly for the stainless steel clad fuel elements, are loaded into the racks, the neutron multiplication 4

factor will be less than the 0.95 limit. Section 3.6 of the Safety Evaluation considered the dropping of a fuel element Or ship-ping cask in the FESW. It was concluded that neither of these events would cause radioactive releases and off-site consequences greater f than the guidelines of 10 CFR Part 100. Technical Specification 4.2.8.5 imposes requirements for items that are handled in or near the FESW. No new technical specifications or changes are contemplated or proposed. In response to the issue of Criterion III, Amendment No. 18 to Pro-visional Operating License No. DPR-45, dated February 4, 1980, con-tained Section 3.3 which concludes the installed fuel racks and pool structures can withstand a dropped heavy load (50-ton cask) and subsequent damage will not be to the extent to adversely affect the leak-tight integrity of the storage well (i.e., will not cause ex-cessive water leakage from the FESW). Technical Specification 2.12.5 specifies the minimum water level above stored irradiated fuel. No new Technical Specifications or changes are contemplated or proposed. b. Where reliance is placed on the operation of the Standby Gas Treatment System, discuss present and/or proposed technical specifications and administrative or physical controla provided to ensure that these assumptions remain valid. DPC RESPONSE: No credit is taken for operation of the LACBWR radioactive gas handling system for any accident involving dropped loads in the reactor vessel or FESW. Where reliance is placed on other site-specific c. considerations (e.g., refueling sequencing), provide present or proposed technical specifications, and discuss administrative or physical controls provided to ensure the validity of such considerations. Y --

Mr. Dennis M. Crutchfield, Chief LAC-8114 Operating Reactors Branch No. 5 February 24, 1982 DPC RESPONSE: No such considerations are considered in any accident analysis, d. Analyses performed to demonstrate compliance uith Criteria I through III should conform to the guide-Lines of NUREG-0612, Appendix A. Justify any exception taken to these guidelines, and provide the specific information requested in Attachment 2, 3, or 4, as appropriate, for each analysis per-formed. DPC RESPONSE: The submittals to the Nuclear Regulatory Commission as listed on page 2 of the Safety Evaluation for Amendment No. 4 to Provisional Operating License No. DPR-45 dated October 22, 1975 with additional submittals for the safety evaluation for Amendment No. 18 to Pro-visional Operating License No. DPR-45, dated February 4, 1980 demonstrate compliance with Criteria I through III. ' 2. 3 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN PLANT AREAS CONTAINING EQUIPMENT REQUIRED FOR REACTOR SHUTDOWN, DECAY HEAT REMOVAL, OR SPENT FUEL POOL COOLING. NUREG-0612, Section 5.1.5, provides guidelines concerning the design and operation of toad-handling systems in the vicinity of equipment or componente required for safe reactor shutdoun and decay heat removal. Information provided in response to this section should be sufficient to demonstrate that adequate measures have been taken to ensure that in these areas, either the likelihood of a load drop uhich might prevent safe reactor shutdoun or prohibit cantinued decay heat removat is extremely amatt, or that damage to such equip-ment from load drops uilt be limited in order not to result in the loss of these safety-related functions. Cranes uhich must be eval-uated in this section have been previously identified in your re-sponse 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 small for att loads to be carried and the basis for this evaluation (i.e., complete compliance uith NUREG-0612, Section 5.1.6, or partial com-pliance supplemented by suitable alternative or additional design features). For each crane so evaluated, provide the load-handling-system (i.e., crane-load-combination) informa-tion specified in Attachment 1. =-. _~ --. l-Mr. Dennis M. Crutchfield, Chief LAC-8114 Operating Reactors Branch No. 5 February 24, 1982 DPC RESPONSE: The crane identified in 2.2-1, the Reactor Containment Building crane, is not designed nor upgraded to be a single-failure-proof load handling system, i.e., is not in complete compliance with NUREG-0612, Section 5.1.6. ) 2. For any cranca identified in 2.1-1 not decignated as single-failure-proof in 2. 3-1, a comprehensive hazard evaluation should be provided which includea the follou-ing information: The presentation in a matrix format of all heavy i a. loads and potential impact areaa uhere damage might occur to safety-related equipment. Heavy loada identification ahould include designation and weight or croac-reference to information pro-vided in 2.1-3-c. Impact areaa should be identi-fied by construction zoneo and elevationa or by come other method cuch that the impact area can be located on the plant general arrangement dravinga. Figure 1 providea a typical matrix. DPC RESPONSE: I For all loads except casks and FCP tabulated under 2.1-3-c, the entire handling operation is done on the 701' level of containment (Figure 6-1). All components for reactor shutdown and decay heat removal, are located below grade floor, elevation 642'9". All components for spent fuel 1 pool cooling are located below 701' level floor, and most are below intermediate floor, elevation 667' (Figure 6.4). Therefore based on the separation between potential impact area and components for reactor shutdown, decay heat cooling, or FESW cooling, for all heavy loads tabulated under 2.1-3-c, there is no likelihood of damage to these systems. However, for any cask handling, there exists a potential for damage, if the cask were dropped in the access hatch area, to the control rod drive lower mechanisms. This potential has been analyzed and protective actions specified (see Safety Evaluation for Amendment No. 4, dated October 22, 1975). With the protective actions and modifications complete, health and safety of the public will not be endangered. 1 p ,y ,.,,..,m, -.m -.,_~.,,,-,,,.,c. _.,,,,,_.___r.,_,_.,y m, r,.yy-,,__re __e_.., x__

Mr. Dennis M. Crutchfield, Chief LAC-8114 Operating Reactors Branch No. 5 February 24, 1982 The cask drop in the access hatch area will not, by reason of separation, impinge on either decay heat or FESW cooling syatems. For the lift of the lA FCP, tabulated under 2.1-3-c, to the 701' level, there exists a potential for damage to decay heat system piping, which is branch connected to the 1A forced circulating loop. No potential for damage exists to either reactor shutdown or FESW cooling systems by reason of separation for a drop of the 1A FCP. The lift of the 1B FCP to the 701' level does not pose a damage potential to either reactor shutdown, decay heat or FESW cooling systems, because it is located in a cubicle remote from these systems. The lift of 1A FCP would only be necessitated by major maintenance. Each of the pumps has been lifted to the 701' level once in 15 years of operation. This frequency, and because only the lift of 1A FCP, imposes a damage potential, is evidence that danger to the health and safety is acceptably low. b. For each interaction identified, indicate uhich of the load and impact area combinationa can be eliminated because of separation and redundancy of safe ty-related equipmen t, mechanical stops and/or electrical interlocka, or other site-specific considerations. Elimination on the basis of the aforementioned consideration should be supplemented by the follouing specific in forma tion : DPC RESPONSE: All load and impact area combinations are eliminated because of separ-ation, except casks and 1A FCP lifts, as detailed in response to 2.3-2-a above. There are no mechanical and/or electrical interlocks or other site-specific considerations used as protective features for reactor systems. Redundancy is extensively used in the reactor shut-down system, but redundancy is limited in the FESW cooling system, and there is no redundancy in the decay heat system itself. (1) For load / target combinatione eliminated because of separation and redundancy of cafety-related equipment, discuss the basis for determining that load dropa citt not affect continued system operation (i.e., the ability of the system to perform its safety-related function). ___ _ _ _ _ _ _ _ _

Mr. Dennis M. Crutchfield, Chief LAC-8114 Operating Reactors Branch No. 5 February 24, 1982 DPC RESPONSE: See response in 2.3-2-b. above. _The alternate means of reactor shut-down,-other than control rod drives, is with the boron injection system. This. system is also separated by at least one floor from any load-impact area. Alternatives in FESW cooling are analyzed and deemed acceptable in the Safety Evaluation to Amendment No. 18 to Provisional Operating License No. DPR-45, dated February 4, 1980. Alternate decay heat removal is available by means of the reactor coolant purification system. The piping for this system is connected to the IB FCP loop, where the decay heat piping is connected to the 1A FCP loop. During the plant conditions, i.e., reactor pressure atmospheric and temperature less than 120'F., when a lift of the 1A FCP would be allowed, the purification system has the capability to remove decay heat. (2) Where mechanical stops or electrical inter-Locks are to be provided, present details shouing the areas uhere crane travet vill be prohibited. Additionally, provide a discussion concerning the procedures that are to be used for authorizing the bypassing of interlocks or removable stops, for verifying that inter-Locks are functional prior to crane use, and for verifying that interlocks are restored to oper-ability after operations which require bypassing have been completed. DPC RESPONSE: See response in 2.3-2-b above. Electrical interlocks are utilized during cask handling on the 701' level floor. No mechanical stops are utilized for any lift. The electrical interlocks and their use are analyzed and deemed acceptable in Safety Evaluation for Amendment No. 4 to Provisional Operating License No. DPR-45, dated October 22, 1975. 3

l Mr. Dennis M. Crutchfield, Chief LAC-8114 Operating Reactors Branch No. 5 February 24, 1982 (3) Where load / target combinations are elimin-ated on the basis of other, site-specific considerations (e.g., maintenance sequencing), provide present and/or proposed technical specifications and discuss administrative procedures or physical constraints invoked to ensure the validity of such considerations. DPC RESPONSE: See response in 2.3-2-b above. No load / target combinations are eliminated on the basis of site-specific considerations. c. For interactions not eliminated by the analysis of 2.3-2-b above, identify any handling systems for specific loads uhich you have evaluated as having sufficient design features to make the likelihood of a load drop extremely small and the basis for this evaluation (i.e., complete compliance with NUREG-OG12, Section 5.1.6, or partial compliance supplemented by suitable alternative or additional design features). For each so evaluated, provide the load-handling-system (i.e., crane-load-combin-ation) information specified in Attachment 1. DPC RESPONSE: All interactions are eliminated in responses to 2.3-2-b, above. d. For interacti:na not eliminated in 2.3-2-b or 2.3-2-c, above, demonstrate using appropriate analysis that damage vould not preclude operation of sufficient equipment to allou the system to perform its safety function follouing a load drop (NUREG-OG12, Secticn 5.1, Criterion IV). For each analysis so conducted, the following information should be provided: (1) An indication of whether or not, for the specific load being investigated, the overhead crane-handling system is designed and constructed such that the hoisting system vill retain its load in the event of seismic accelerations equivalent to those of a safe shutdoun earthquake (SSE). _ _ _ _ _ _ _ _ _ _ _ _ _ - _

Mr. Dennis M. Crutchfield, Chief LAC-8114 Operating Reactors Branch No. 5 February 24, 1982 (2) The basis for any exceptions taken to the analytical guidelines of RUSEG-0612, Appendix A. (3) The information requested in Attachment 4. DPC RESPONSE: All interactions are eliminated in responses to 2.3-2-b, above.

SUMMARY

It is recognized by LACBWR that the Reactor Containment Building polar crane handling equipment could fail during any lifting event. There are tests and inspections on the crane which are performed in accordance with ANSI B30.2.0-1976. This, in conjunction with the crane preventive maintenance program, minimizes the potential for load drops. No credit is taken for the tests in any load drop analysis, because protection from a load drop rather than prevention of a drop has been developed. If you have any questions, please contact us. Very truly yours, DAIRYLAND POWER COOPERATIVE V } A Ah C Frank Linder, General Manager FL: HAT:af cc: James G. Keppler, Reg. Dir., NRC-DRO III NRC Resident Inspector ll h 1:r'_t:;w n. ,s l x,). - - - -.. ', .-I. Ii!IWe : l jqt@} l ~ ~~ =.. _ y MN. fg'N / h j ~ M M ) 6 ,. 7.,7 - ; --sy -- - N .meh ,7 7 h t. I m ~ . A A7 1' -T iis _._ _ A \\w m w = pI-=,w:. x i-l1 \\w{\\f~ly.-l{ ~~ ?,, b} ~4. ' .-\\ .: ig ' g/, f I h- 'N.s, .,- / t ~ d 72-f 1,AiN; & 3\\ l \\ (p V ,/ </. i l m.w. y. u... ,y r 4 - --. w / w l ' ' g3. _ _ s N , - r = --- y s N \\ ^ N,N__ _ _ _7P-- f e d s-j e \\ x $V. N ' N- /j f% ';3 )x & @s l

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