ML20010H551

Forwards Response to NRC 801222 Request for Addl Info Re Control of Heavy Loads

ML20010H551
Person / Time
Site:	Zimmer
Issue date:	09/22/1981
From:	Borgmann E CINCINNATI GAS & ELECTRIC CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-0612, RTR-NUREG-612 NUDOCS 8109250228
Download: ML20010H551 (73)

Text

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e THE CINCINNNI'I GAS & ELECTRIC COMPANY CINCINN ATI. OHIO 45201 6i, Q

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SENtOR veCE PRESIDENT f Up ,-

(h %SEp,i Docket No. 50-358 1S87 % d Septen.ber 22, 1981

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Mr. Harold Denton, Director (6',/

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Office U.S. Nuclearof Nuclear Reactor Regulatory RegulationN@/,r7pV<P/

Commission k Washington, D.C. 20555 RE: WM. H. ZIMMER Nut, LEAR POWER STATION -

UNIT 1 - SUPPLEMENTAL INFORMATION IN RESPONSE TO NRC LETTER OF DECEMBER 22, 1980 REGARDING CONTROL 0F HEAVY LOADS

Dear Mr. Denton:

In reply to the NRC letter of December 22, 1980 from Darrell G.

Eisenhut to all licensees of operating plants and applicants for operating licenses and holders of construction permits, there are attached eight copies of supplemental information in res;;onse to Sections 2.2 and 2.3 of Enclosure 3 to the above referenced NRC letter. Our response to Section 2.1 of Enclosure 3 to the above referenced NRC letter was submitted on June 24, 1981.

Very truly yours, THE CINCINNATI GAS & ELECTRIC COMPANY E. A. BORGMANN EAB: dew Enclosure

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5 cc: Without Enclosure John H. Frye III David K. Martin M. Stanley Livingston Frank F. Hooper George E. Pattison Andrew B. Dennison

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Troy B. Conner, Jr.

James P. Fenstermaker State of Ohio )

Steven G. Smith County of Hamilton)ss William J. Moran J. Robert Newlin Sworn to and subscribed before me this Samuel H. Porter JM day of September,1981.

James D. Flynn W. F. Christianson W. Peter Heile James H. Feldman, Jr.

John D. Woliver

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Mary Reder f M LYN J.LUERSEN 8109250228 810922 Nctry M!ic. State cf Ohio

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i DOCKET NO. 50-358 WM. H. ZIMMER NUCLEAR POWER STATION UNIT 1 i . CONTROL OF HEAVY LOADS September 22, 1981 I

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DOCKET NO. 50-358 WM. H. ZIMMER NUCLEAR POWER STATION

,UMJT 1 CONTROT,3;_ ? .'.VY LO ADE SA : - or 22, 1981 In response to DarreILG. Eisenhut's Decembet 22, 1980 letter con-cerning control of heavy loads we submit the following information:

Question 2.2.1 Identify by name, type, capacity and equipment designator, any cranes physically capable (i.e.,

ignoring interlocks, movable mechanical stops, or operating procedures) of carrying loads over spent fuel in the storage pool or in the reactor vessel.

Resnonse 2.2.1 The following cranes are physically capable of carrying ,

loads over the spent fuel in the storage pool or in the reactor vessel.

Table 1 Equipment .

Item No. No. Name Tm Capacity 101 1HC01G Main Reactor Bridge Crane 110 Ton Main Bridge Crane Hook and 10 Ton Auxiliary Hook 102 1HCO2RB Fuel Handling Jib Crane 1 Fuel Jib Crane Assembly and handling tool 103 1HCO3RB Channel Hand- Jib Crane 200 lbs.

l ling Boon Jib Crane .

l 104 1HC04RB Fuel Handling Jib Crane 1 Fuel Jib Crane Assembly and handling tool 113 IHC13RB Refueling Bridge Crane 1 Fuel l Platform Assembly and handling tool l

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'115 1HC15RB Service Plat- Jib Crane 1 Fuel form Jib Crane Assembly and handling tool Question 2.2.2 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 where, following any failur~e, such a load may drop into the reactor vessel or spent fuel storage pool.

Response

2.2.2 Heavy loads are defined as loads greater than 1 fuel assembly plus tho veight of the handling tool. Based on the above heavy i)ad definition all cranes on the refueling floor except for the main reactor building bridge crane, 1HC01'3, can be excluded from review. ,

Question 2.2.3 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 evaluation (i.e., complete compliance with NUREG 0612, Section 5.1.6 or partial compliance supplemented by suitable alternative or additional design features).

For each crane evaluated, provide the load handling system (i.e. , crane loading combination) information specifying in Attachment 1.

Response

2.2.3 The main reactor building bridge crane main hook (110 ton) (equipment no. 1HC01G).has been reviewed in de-tail in the Safety Evaluation Report (SER) and accepted.

Section 9.1.4, Fuel Handling Systems (pages 943 and 9-4) of the SER addressed tha bridge crane. The crane is described in detail in ESAR Section 9.1.4.2.2. In

• addition, as stated in the SER, it was concluded in the SER that the fuel handling system is designed to safely handle fuel assemblies- from receipt cf new fuel to shipping fuel. The load handling system for the reactor building bridge crane main hook are listed in

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Table 2 which has been revised from the table previously submitted.

The narrative contained in the Mm. H. Zimmer Nuclear Power Station FSAR, Section 9.1.4.2.2, describes the design features incorpor-ated in the 110 ton bridge crane. These features provide reason-able assurance that safe handling of heavy loads on the plant refueling elevation is accomplished.

The load handling systems listed in Table 2 were selected to meet the requirenents of NUREG 0612 Section 5.1.6, (1), (b), -

(i) , or (ii) 1.e. redundant slings are provided such that a single component failure or malfunction in the -sling will not result in uncontrolled lowering of the load, or in selecting the sling, the load used will be twice the static load required.

The adequacy of interfacing lift points for refueling floor loads are under review by our architect engineers. Equipment design changes will be accomplished if necessary. Changes to equipment such as the reactor vessel head or vessel intermals would only be completed if added safty warrants any possible deleterious effects to the components. The shactic selected for lifting the spent fuel pool plugs was limited by physical dimension constraints and only provides a safety factor of 6.5 to ultimate breaking strength. Any changes made to the interfacing lift point would of course proride for use of equipment of a

- greater rating.

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Question 2.2.4 The crancs identified in 2.2.1, above, not categorized according to 2.2.3, demonstrate the.t the criteria of NUREG 0512 Section 5.1 are satisfied. Compliance with criteria IV will be demonstrated in response to Section 2.4 of this request. With respect to criteria I thru III, provide a discussion of your evaluation of determination of compliance.

Response

2.2.4 The following cranes are addressed in 2.2.4 which are not excluded by 2.2.3.

M-19 Sheet 12 Equipment N-Item No. No. Name Type Capacity 101 1HC01G Main Reactor Bridge Crane 10 ton Building Bridge Auxiliary Hook Crane (Aux Hook only) 192 1HCO2RB Fuel Handling Jib Crane. 1 Fuel Jib Crane Assembly and i

handling tool 103 1HCO3RB Channel Hssd- Jib Crane 200 lbs.

ling Boon Jin ,

j Grane 104 IHC04RB Fuel Handling Jib Crane 1 Fuel Jib Crane Assembly and handling tool

113 1HC13RB Refueling Bridge Crane 1 Fuel Assembly and Pictiorm handling tool

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115 1HC15RB Service Plat- Jib Crane 1 Fuel Assembly l

form Jib Crane and handling tool The Reactor Building Bridge Crane auxiliary hook (10 tons) does not meet the single failure criteria. The use of this crane is limited to a maximum load of 1 fuel assembly and its handling tool when operating over the spent fuel pool. The 10 ton

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l auxiliary hook is also used for hoisting new fuci, replacement control rods, fuel channels and incore detector strings from the equipment access building to the refueling floor. The auxiliary book is utilized during new fuel inspection for the transport of individual shipping crates and fuel assemblies during the fuel inspection process.

l In addition to the above loads, the 10 ton auxiliary hook is used to move loads which are stored along the west wall and can not be reached with the main hook. Lateral movements of these loads shall be kept to the minimum required for attach- '

ment to the main hook.

The jib cranes and the refueling platform operate over the spent fuel and the vessel. If a postulated accidental drop of a load crane caused damage to the spent fuel, release of radioactive materials shall be contained as addressed in the FSAR. The Fuel Pool Ventilation Erhaust Plenum Radiation Monitoring Subsystem (See FSAR Section 7.1.2.1.1.1,7.2 and 7.6.2.3.7) will initiate control signals in the event the radiation level exceeds a predetermined level to isolate the fuel pool vent syst _, to initiate the standby gas treatment system, and to close containment purge and vent valves. The redundancy and arrangement of channels assure that no single failure can prevent isolation when required. During refueling operation, the monitoring system acts as an engineered safe-

. guard against the consequences of a refueling accident or the rod drop accident. The above actions will assure the 10CFR100 limits are met. In addition, the main Plant Vent Stack Radiation Monitoring subsystem (see FSAR Section 7.1.2.11.6) monitors the radioactivity within the main plant stack to generate alarms <

if the activity level reaches either short term or long term release limits.

The Reactor Building Ventilation and PIessure Control System l (see FSAR Section 7.1.2.1.14) is designed to hold th2 Reactor Building pressure at a negative pressure of 1/4" H 2 O gauge

under all normal operating conditions. If ' radioactivity is detected in the exhaust gas from the building, the control system isolates the building and directs the ventilation exhaust to the standby gas treatment system. The standby gas treatnant system control and instrumentation (see FSAR 7.1.2.1.27) are designed to meet the following safety designs bases

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Start the standby gas treatment system to maintain the reactor building at a negative pressure to assure infil-tration and to filter the radioactive particulates and l

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iodine from the influents in the case of a loss of coolant accident or fuel handling accident.

b. The standby gas treatment. system will respond automatically so that no action is required of station operators follow-ing a loss of coolant accident or fuel handling accident.

c. The responses of the standby gas treatment system will be indicated on the main control board.

d. No single failure, maint'enance, calibration or test .

operation will prevent operation of the standby gas treat-ment system.

c. And the physical event accompanying a loss of coolant or fuel handling accident will not prevent correct functioning of the standby gas treatment system controls and instru-mentation.

The draft Technical Specifications shown in Table 3 serve to ensure ope rability of the standby gas treatment system and it's initiat i; . instrumentation.

A Spent Fuel Pool Leak Detection system to monitor leakage from the fuel pool liner and seal bellows is provided to activate a.:

annunciator in the event of a system leak of sufficient mag-

.nitude (see FSAR Section 7.6.1.6.9). Flow switches are located in the fuel pool channel drain and in the fuel pool bellows seal drain. A main control room alarm is activated when leakage reaches this predetermined value. .

The fuel handling accident is addressed in Section 15.1.41 of the FSAR where the most severe accident from a radiological viewpoint is addressed. Description of the accident, operator actions, methods, assumptions and conditions are addressed in the referenced section. The results and consequences are covered in FSAR Section 15.1.41.5.1.2. .

The spent fuel pool storage racks are designed to meet seismic Category 1 requirements and to withstand the impact resulting from a falling weight possessing 2000 ft-lb. kinetic energy.

When subjected to this impact, those members which maintain spacing to assure keff less than or equal to 0.95 remain intact.

Load movement paths are provided to avoid travel over the spent fuel pool or the reactor well. The paths shall be as direct as practical and preplanned as covered in plant implementing procedures.

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The main reactor building bridge crane has electrical interlocks provided to prevent movement of heavy loads above fuel in the sacnt fuel pool rods. The crane operator can override and enter tae interlock protected area when approved by the maintenance supervisor or abift supervisor. The keys for the interlock control bypass are controlled by the shift supervisor. The refueling floor opcrating procedures specifically cover the use of the override and the authorized reasons to enter the area. Overriding the interlocks shall o'nly be done in accordance with plant implementing procedures. Movement of the fuel pool gates and shipping cask pit gate require bypassing the fuel pool interlocks. Movement of these gates shall bc accomplished with the single failure proof 110 ton hook with rigging rated at0>

two times the gate's weight.

Attached is the M-17 drawing showing the heavy load movement paths. Each heavy load shall have specific movement path selected to minimize the consequences of a load drop. Any deviation to the specified movement path shall be approved by

! the maintenance engineer prier to lifting the load. Crane operators shall be instructed to minimize the lift of any loads to as low as practicable.

Question -

2.3.1 Identify any cranes listed in 2.1-1, above, which you

. have evaluated as having significant design features to make the likelihood of a load orop extremely small for all loads to be carried and the basis for this evaluation (i.e., complete. compliance with NUREG 0612, Section 5.1.6, or partial compliance supplemented by suitable alternative or additional design features).

For each crane so evaluated, provide the load handling-system (i.e., crane load combination) information specified in Attachment 1.

Response

The Reactor Building Bridge Crane is the only single

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failure proof crane at the plant site. See Response 2.2.3.

Question 2.3.2 For cranes identified in 2.1-1 not designated as single failure proof in 2.3-1, a comprehensive hazard evalua-tion should be provided which includes the following l

information:

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Question 2.3.2a. The Presentation in a matrix fc..nat of all heavy loads and potential impact areas where damage may occur to safety related equipment. Heavy loads identification should include designation and weight or cross reference to information provided in 2.1-3c. Impact areas should be identified by construction zone: and elevations by some other method such that the impact area can be

located on the plant general arrangement drawings.

Response

2.3.2a Attached is a matrix for all cranes (lifting loads greater than 1 fuel assembly plus handling tool) listed in Table 1 submitted in the response to Section 2.1 of .

Enclosure 3. The matrix references M-19 series of drcwings which are the equipment removal drawings for the plant. Those drawings are based on general arrange-ment drawings and show equipment access paths. These matrices are included as attachment 1 to this'. letter.

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t Question 2.3.2b For each interaction identified, indicate which load and impact area combinations can be eliminated be-cause of separation and redundancy of safety related equipment, mechanical stops and/or electrical inter-locks, or other site specific considerations:

Response

2.3.2b The following equipment can be eliminated by one of the above reasons:

Item 101 Reactor Building Bridge Crane Main Hook (110T). The main hook has been addressed in Response 2.2.3. The Auxiliary Hook (10T) shall be administrative 1y controlled to lifting item loads less than I fuel assembly (plus weight handling tool) over the spout fuel pool and to reach items on the west side of the

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reactor building and refueling finor where the main hook cannot reach. These items shall be lifted and moved a minimum distance to allow the main hook to be used. The auxiliary hook is also used, with the crane interlocks operable, to hoist new fuel shipping crates, fuel channel chippi.ig crates, replacement control rod blade shipping crates and incore detector shipping crates to and from the equipment access building and the plant refueling floor. The auxiliary hook is utilized for movement of new fuel shipping crates during inspection of new fuel.

Item 107 RHR and RBCCW (IB) Heat Exchanger, Item 197, is a 20 l

ton monorail overhead hoist to be used for tube bundle removal and overhaul of the IB RRCCW Heat Exchanger I and RHR Heat Exchanger 1A and 1B. Sufficient separ-l ation exists insuring that inadvertant drop of any of the above components would not cause damcge to any other system required for safe shutdown or decay heat removal. The IB RBCCW Heat Exchanger is separated by

, 12 ft. from the north bank hydraulic control units

! and by two floors from the RHR Heat Exchankers. The i RHR Heat Exchangers are located in separate cubicles and are located 2 floors below the 1B RBCCN Heat Exchanger.

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Item 108 Main Steam Hatch Slabs and Isolation Valves I Administrative controls shall be applied to assure that the main steam hatch slabs are not removel during plant operation. Inadvertant dropping of the j main steam hatch slabs after cold shutdown will not l effect plant safety. F1milarly, inadvertent dropping of any of the main stes, isolation valve components j or feedwater val /e components, after they have been i released for maintenance, will not have any effect on plant safety or decay heat removal.

Item 111 Hotch. Slabs and RCIC Liaintenance Panel H22-P022 contains one steam line flow switch for each main steam line and the recirculation loop flow transmitters feeding the B flow unit for APRM flow biased scrams. Based upon single failure proof i criteria employed in the design of these systems j their failure can neither cause nor prevent the com- ,

pletica of a safety function. l Item 112 RBCCW 18 Heat Exchanger 1A I Item 11'2 is a 20 ton monorail overhead hoist to be used for tube bundle removal and overhaul of the 1A '

, RBCCW Heat Exchanger. Sufficient separation exists to ensure inadvertant drop of the heat exchanger would not cause damage to any other system required for safe shutdown or decay heat remo'ral.

Item 118 Low Pressure Core Spray The inadvertant drop of a low pressure core spray pump component, has a very small probability of damaging the RHRA pump as evidenced by equipment separation. Even the assumed total loss of ECCS division I has no effect upon safe shutdown and decay heat removal since two redundant divisions of ECCS remain operable.

Item 119 RHR Pumps The inadvertant drop of any RHR pump component, has

. a very small probability of damaging the RHR pump or LPGS pump in it's room as evidenced by equipment physical separation. Even the assumed total loss of

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