ML19344A816
| ML19344A816 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry, Turkey Point  |
| Issue date: | 03/27/1980 |
| From: | Overbeck G FRANKLIN INSTITUTE |
| To: | George H Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19344A817 | List: |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, TASK-A-36, TASK-OR NUDOCS 8008220275 | |
| Download: ML19344A816 (6) | |
Text
O THIS DOCUMEf1T C0f1TAlf1S POOR QUAUTY PAGES INFORMAL TECHNICAL COMMUNICATION i
Date M acu O,
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From:
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Os cut s e e,c QO United States Nuclear Regulatory Conmission Franklin Research Center Washington, D.C. 20555 Philadelphia, PA 19103 Attention:
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Reference:
FRC Project C5257 NRC Contract NRC-03-79-ll8 NRC TAC No.
o7914 FRC Task
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. Attachment (1) 2 2
1 REQUEST FOR ADDITIONAL INFORMATION BROWNS FERRY UNITS 1, 2, AND 3 REACTOR BUILDING CRANE 1.
Tables 12.2-14'and 12.2-15 of the Browns Ferry FSAR provide the load combinations and allowable stresses for the reactor building crane.
Indicate the degree to which these allowable stresses comply with those provided in CHAA Specification No. 70-1975 for welded box girders, bridge end trucks, and trolley frames.
2.
In response to R. A. Purple's letter of June 10, 1975, to James E. Watson, which requested a documented comparison of the Browns Ferry Nuclear Plant reactor building crane with the positions given in Branch Technical Position APCSB 9-1, an evalua-tion was provided for Position 3.o which is not clear.
Describe the features that are inherent in the maxspeed D.C. adjustable voltage systems used on both hoist and travel drives to prevent abrupt change in motion.
Describe whether or not the electric power system allows the brakes to be released while the hoist or drive motors are not energized.
3.
In response to Position 4.c of the Branch Technical Position APCSB 9-1, a statement was made that the maximum-working-load capacity will be maintained at 100% of the design-rated load (DRL). A single-failure-proof crane should be designed to handle the maximum critical load (MCL) that will be imposed.
A single-failure-proof crane may handle non-critical loads of greater magnitude th'an the MCL.
For such cases the maximum non-critical load or maximum working load can be the design-rated load.
For the reactor building crane, define the MCL.
If the MCL is within 15% of the DRL, define any degrcdation considered for components (due to wear and exposure) and describe the procedures that are followed to detect and correct degraded conditions.
Describe any design features which will effectively limit the load which the wire rope and other wear-susceptible components will experience.
4.
NUREG-0544 entitled, " Single-Failure-Proof Cranes for Nuclear Power Plants," states that cast iron should not be used for load-bearing components.
Indicate whether or not cast iron was used for load-bearing components.
5.
In addition to the discussion provided in FSAR Section 12.2.2.5.1, describe the seismic analysis employed to demonstrate that the reactor building crane can retain.
v.
the MCL during a seismic event equal to'a safe shutdown earthquake and provide a description of' the method of analysis and the assumptions used. The description of.the method of. analysis should include a discussion of the analytical model.
The description of assumptions should include the basis for selection of trolley and load position.
6.
Lamellar tearing is a ductile failure of the parent steel and occurs while the steel is cooling after the welding process.
The tearing serves to relieve the tensile stresses imparted during the welding process.
Such stresses, and the subsequent tearing, can be directly related to the welding method (heat imput),
the difference in strength between base and weld metals (weld metal is typically much stronger), the amount of restraint of the welded joint, the geometry (con-figuration) of the joint. itself, and other factors.
Identify any fabrication requirements, design requirements, and manufacturing / assembly procedures employed to minimize the introduction of lamellar tearing.
7.
In response to Position 1.c of the Branch Technical Position APCSB 9-1, a reference was made to FSAR Section 12.2.2.5.3.
In keeping with a " defense-in-depth" philoso-phy, reliance on single failure protection for critical parts is not acceptable.
For critical rotating parts which must endure a higher number of cycles or varying reversed stresses, clarify that these parts are designed to accommodate cumulative damage from fatigue.
Identify load-bearing rotating parts which are subjected to stress cycles.
This identification should include the material used and the basis for determining satisfactory component design.
8.
In response to Position 2.b of the Branch Technical Position APCSB 9-1, a reference j
was made to FSAR Section 12.2.2.5.2 for details of compliance. FSAR Section l
12.2.2.5.2 does not clearly indicate that the auxiliary handling system of the reactor building crane is single-failure-proof.
Indicate whether or not the aux-iliary hoieting system is employed to lift or assist in handling critical loads (defined in NUREG 0554).
If the auxiliary handling system is employed, describe the degree of redundancy provided for the hoisting system.
9..
For the MCL identified in 3, above, provide the maximum stress (including static and inertia forces) <xt each individual wire rope in the dual reeving system and compare it to the manufacturer's published breaking strength.
If the auxiliary hoisting system is employed to lift or assist in handling critical loads, provide the weight =of the load and the maximum stress in the rope.
Compare this stress to the manufacturer's published breaking strength.
10.
Provide the ratio of drum diameter to the rope diameter for the reactor building crane main hoist.
If the auxiliary hoisting system is employed to lift or assist in handling critical loads, compare the drum and sheave diameters to the rope diameter.
11.
For the_ load-block assembly illustrated in FSAR Figure 12.2-22d, indicate whether or not the attachment points for the safety cables can support a load which is three times the weight of-the load being handled without permanent deformation of any part of the load-block assembly and fuel-cask yC e, other than localized strain concentration in areas for which additional material has been provided.
12.
For the safety cables, provide the factor of safety as compared to the critical load being handled.
Indicate whether or not the safety cables are loaded during fuel-cask-handling operations.
If they are not loaded, describe how the impact loa' ding from a ' hook failure was utilized in establishing the cable design.
s 13.
For other critical loads (assuming spent fuel cask is a critical load), describe the means employed to achieve the equivalent of dual attachment points.
In lieu of dual attachment points, a single attachment point would be acceptable if the factor of safety is increased to ten.
If dual attachment points are not provided for other critical loads, deminstrate that the stress levels are less than 10%
l of the load-carrying capabilit y of the hook.
14.
Since the safety cables are acting in conjunction with the load-handling hook, state whether or not a 200% static-type-load test has been performed on the safety cables.
15.
For the reactor building crane load blocks, indicate whether or not they have been non-destructively examined by surface and volumetric techniques.
- 16..If the auxiliary hoisting system is employed to lift or assist in handling critical loads, describe the extent to which the design is protected against two-blocking, as delineated in NUREG 0554, Item 4.5. -
0 17.
For the lifting devices, identified in your response to Position 3.b of the Branch Technical Position APCSB 9-1, which handle critical loads over the open reactor vessel and spent fuel,-describe the degree of compliance with the following
- criteria, Spec-lal lifting devices shculd satisify the guidelines of a.
ANSI N14.6-1978, " Standard for Special Lifting Devices for for Shipping Containers Weighing 10,000 pounds (4500 kg) or More for Nuclear Materials." This standard should apply to all special lifting devices which carry critical loads as defined by NUREG 0554.
For operating plants, certain inspections and load tests may be accepted in lieu of cer-tain material requirements in the standard.
In addition, the stress design factor stated in Section 3.2.1.1 of ANS1 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.1 This is in lieu of the guideline in Section 3.2.1.1 which bases the stress design factor on only the weight (sta-tic load) of the load and of the intervening components of the special handling device.
b.
Lifting devices that are not specially designed should be installed and used in accordance with the guidelines of ANSI B30.9-1971, " Slings." However, the safety factor should be the ratio between the breaking strength and the maximum combi 6ed static and dynamic load.1 The rating identified on the sling should be in terms of the " static load" which pro-duces 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 which they j
may be used.
18.
Describe the extent of compliance to NUREG 0554 Item 4.7 concerning wire rope protection.
19.
In response to Position 3.h of Branch Technical Position APCSB 9-1, a reference was made to FSAR Section 12.2.2.5.2 which describes safety features which meet the " intent" of this position.
Although the referenced FSAR section does describe safety devices which mitigate the effects of overpower and overspeed, there is no clear indication that these controls are capable of stopping the hoisting movement s! thin amounts of movement such that damage would not occur. Provide an evalua-tion of the controls on the Browns Ferry reactor building crane which limit hoist movement upon sensing an overpower or overspeed condition.
I or the purpose of determining the safety factor, loads imposed by the safe F
shutdown earthquake need not he included in the dynamic loads imposed on the sling or lifting device. _
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- 20. 'For the reactor building crane which employs more than one control stat ion, indi-
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,cate whether-.or not electrical interlocks have been provided to permit only one control station to be operable at any one time.
-21.
In lieu of two-block testing of the main hoist, as delineated in Position 4.b of Branch Technical Position APCSB 9-1, two independent travel-limit switches are
< acceptable provided there is periodic verification of the proper functioning of.
these switches.
Describe the method and frequency of testing of the limit switches.
In addition, the testing requirements of the current limiting device on the -hoist motor, should be discussed as well as any other device which may be required to function in order to mitigate the effects of load hangup.
22.
The operating manual for the reactor building crane should contain, as a minimum, the information described in NUREG 0554.
Indicate the extent to which the infor-mation is contained in the reactor building crane operating ranual.
In addition, verify that the operating requirements for all travel movements incorporated in the design are clearly defined in the operating manual for hoisting and for trolley and bridge travel.
23.
For the reactor building crane, - describe the quality-assurance program used in theLdesign, fabrication, installation, testing, and operation. The program should address the qualification requirements for crane operators.
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