ML20042C005
| ML20042C005 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 03/15/1982 |
| From: | Clayton F ALABAMA POWER CO. |
| To: | Eisenhut 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 NUDOCS 8203290513 | |
| Download: ML20042C005 (12) | |
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Milling Addrxs Alabama Power Company 600 North 18th Street Post Office Box 2641 hrmingham. Alabama 35291 13lephone 205 7834,081 F. L. Clayton, Jr.
Senior Vice President ghgg]3 pgg Ft.ntridge Bwiding itn?.r;outhern eiwtrc system Ma rch 15,1982 rs
\\o Docket Nos. 50-348 50-364 fmcgpp<.<,,
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Director, Nuclear Reactor Licensing 22 g%ack4 rear U. S. Nuclear Regulatory Commission E
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Washington, D. C.
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Attention:
Mr. D. G. Eisenhut fs
Reference:
(1)
May 15, 1981, Alabama Power Company's Interim Actions Response foi Control of Heavy Loads (2)
June 24, 1981, Alabama Power Company's General Requirements for Overhead Handling Systems Response to NRC (3)
TER-C5257-76/443, October 23, 1981, Franklin Research Institute Technical Evaluation Report JOSEPH M. FARLEY NUCLEAR PLANT - UNITS 1 AND 2 NUREG-0612 - CONTROL OF HEAVY LOADS
Dear Mr. Eisenhut:
As requested by members of the NRC Staff, Alabama Power Company hereby submits the attached response to the above referenced Franklin Research Institute (FRI) Technical Evaluation Report.
This response is intended to complement the information provided in our previous submittals (references (1) and (2)).
Also attached are five (5) sets of the safe load path drawings as requested by FRI.
Yours very truly, f*'
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ATTACHMENT NUREG-0612 - CONTROL OF HEAVY LOADS FRC Position:
l 2.1.1.
c.
FRC Conclusions and Recommendations l
On the basis of FRC's review of the Licensee's response, the following cranes should be evaluated for compliance with the general guidelines of NUREG-0612, Section 5.1.1:
1.
drumming station bridge crane I
2.
auxiliary building equipment hatch monorail hoist i
3.
decontamination roor monorial hoist 1
1 4.
blowdown drum storage area bridge crane 5.
spent fuel pool bridge crane
)
6.
diesel generator building hoists 7.
external portable maintenance cranes.
l f
Reponsg:
A.
Cranes 1 - 4 have been evaluated to ascertain whether or not
+
the crane is physically capable of lifting a heavy lead over i
safety-related equipment.
It has been determined that no safety-related equipment is installed under these cranes, or at i
j lower floor elevations, therefore, the guidelines do not apply.
l l
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B.
. Crane 5 is addressed in the response to FRC paragraphs 2.1.2C 1
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and 2.2.1 as this crane only transports loads over spent fuel.
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i C.
Crane 6 is a monorail hoist in the diesel building, therefore, i
only a single load pathway is possible.
Furthermore, the l
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Response _(cont'd[
Page 2 equipment separation in the diesel building assures that no heavy load can damage two trains of safety-related equipment.
Franklin Research Institute stated in Section 2.1.lb of their diesel generator hoist evaluation that " suitable precautions should be added to any existing procedures identifying the heavy loads and their respective handling systems, or the Licensee should physically mark the component with a suitable warning."
The Farley Nuclear Plant diesel generator hoists are mounted upon a monorail sy s t em.
The placement of the fixed rails precludes the lif ting of a heavy load over more than one train of safety-related equipment.
Since the monorails are limited to only one path, identification of heavy loads is u nneces sa ry.
D.
The external portable maintenance cranes were addressed in Alabama Power Company's June 24, 1981 submittal.
These cranes are used outside the plant buildings, thus the specification cf load pathways is not warranted.
FRC Position:
2.1.2 c.
FRC Conclusion The Farley Nuclear Plant does not comply with the criteria of Guideline 1.
In order to satisfactorily comply, the licensee should perform the following:
1.
Verify that safe load paths which have been l
FRC Position (cont'd)_
Page 3 4
f developed in the containment have been incorporated in procedures and equipment layout drawings.
2.
Develop safe load paths for the movement of heavy loads in the auxiliary building which satisfy the criteria of Guideline 1.
3.
Verify that safe load paths will not cause potential damage to irradiated fuel stored in the spent fuel pool.
4 Verify that safe load paths are clearly marked to identify safe load pathways, including any necessary provisions in plant procedures to ensure that any temporary coverings placed over i
permanent markings are suitably marked.
t 5.
Develop the necessary administrative procedures to verify that deviations from established load pathways required written alternatives which are specifically approved by the plant safety review committee.
Response
A.
The safe load pathways for heavy loads inside containment have been incorporated into the appropriate plant procedures.
These procedures contain drawings that define the safe load pathways.
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Response (cont'd)
Page 4 B.
All cranes in the auxiliary building except the demineralizer hatch monorail hoist are prevented f rom carrying a heavy load over safety shutdown equipment by physical separation.
In removing one of the demineralizer hatch covers there is a possibility that the demineralizer hatch monorail hoist could indirectly drop the hatch cover onto portions of piping associated with one train of the boric acid transfer system.
Since, at most, only one train can be disabled by a single drop of the hatch cover, loss of the required safe shutdown functions will not occur and safe load pathways are not required.
C.
Alabama Power Company fuel handling procedures administratively prevent heavy loads from being transported over spent fuel in the storage pool.
The only heavy load handled by the spent fuel pool bridge crane is the spent fuel pool transfer slot gate which weights approximately 3000 lbs.
At the beginning of j
fuel transfer operations the transfer slot gate is moved from its normal position directly to its storage position which is located immediately adjacent to its normal position.
This procedure is reversed at the end of fuel transfer operations.
This operation does not move the transfer slot gate into a I
position where it could cause damage to the stored fuel.
Alabama Power Company is currently in the process of licensing new spent fuel racks for Unit 2 and will submit proposed Technical Specification changes which would allow reracking of I
the Unit 1 Pool.
These new racks are designed such that the
Response (cont'd)
Page 5 impact of the transfer slot gate on top of the racks during normal movement of the gate will not result in damage to spent i
f uel assemblies.
The above existing controls which preclude travel of heavy loads over spent fuel will remain in effect until af ter the new racks are installed.
New racks are scheduled to be installed in Unit 2 prior to the first ref ueling outage and af ter the fourth refueling outage for Unit 1.
D.
The safe load pathways inside containment are clearly defined by drawings in the plant procedures.
The marking of the floor would provide no useful information to the crane operator and would merely add an additional burden to the work load of the personnel invol ved in plant maintenance operations and could increase the duration of outages.
The crane operator's training includes specific instructions concerning the handling of loads within the defined safe load pathways inside containment.
The pathways are generally defined by large pieces of equipment and, as such, are clear to the operator without physically marking the floor.
E.
Any deviation from plant procedures, including deviations from safe load pathways, are reviewed and approved as required by plant Technical Specifications.
The FNP Technical Specifications permit deviations from approved procedures under certain conditions.
The deviation from a safe load pathway, while permitted by the Technical Specifications as a procedural deviation, is subject to the same restrictions as deviations from any other plant procedures.
Page 6 FRC Position:
2.1.3 c.
FRC Conclusion Farley Unit 1 complies with Guideline 2 in that procedures are being developed by the Licensee to meet the criteria of this guideline.
When completed, these proc?dures should be readily available for review by the NRC Staff.
Response
These procedural modifications have been completed and are available for review by NRC I&E Inspe: tors.
FRC Position 2.1.4 c.
FRC Conclusions and Recommendations The Farley Nuclear Plant does not comply with Guideline 3.
To satisfactorily comply, the Licensee should implement crane operator programs which satisfy the requirements for training, yualification, and conduct specified in Chapter 2-3 of ANSI B30.2-1976.
Response
The requirements of ANSI B30.2-1976 Chapter 2-3 have been incorporated into the crane operator training, qualification, and conduct procedures.
FRC Position:
2.1.5 c.
FRC Conslusions and Recommendations The Farley Nuclear Plant does not comply with Guideline 4.
To satisfy the requirements of this guideline, the i
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FRC Position: (cont'd)
Page 7 i
Licensee should perform a point-by-point comparison of all special lifting devices in use with the applicable criteria of ANSI N14.6-1978 and the stress design factor specified in the guideline.
Response
{
Two special lifting devices have been identified at Farley Nuclear i
Plant:
Reactor Vessel head lifting device Core upper internals lifting device A review is under way to determine if any other rigging devices should be classified as "special lifting devices."
The comparison of the' lifting devices with ANSI N14.6-1978 has been initiated.
The results of these studies is scheduled for submittal by June 1, 1982.
FRC Position:
2.1.6 c.
FRC Conclusions and Recommendations The Farley Nuclear Plant does not comply with Guideline S.
In order to comply, the Licensee should perform the following:
j 1.
Verify that slings are installed and used in accordance with ANSI B30.9-1971.
2Property "ANSI code" (as page type) with input value "ANSI B30.9-1971.</br></br>2" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..
Verify that the load used in selecting and marking the proper sling is based upon the sum j
of the maximum static and maximum dynamic loads.
3.
Verify that slings which are restricted in use to certain cranes are clearly marked as to the I
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FRC Position (cont'd)
Pags 8 cranes with which they may be used.
Verify 4
hook speeds for the polar, spent fuel, and cask cranes constitute a slow speed such that rigging will not be overstressed due to dynamic loading.
Response
The slings used for rigging loads at FNP are installed and used per ANSI B30.9-1971.
The ANSI standard was developed for lifting loads under dynamic conditions.
Load limits for slings under ANSI B30.9-1971 are mass handling limits, not dead weight stress limits.
The dynamic component is included in the standard through the application of significant safety factors.
It is, therefore, inappropriate to apply additional dynamic stress criteria in the sling selection progress.
The slings are developed for the load, not for a specific crane; therefore, the markings on slings contain load limits rather than a designation of the specific cranes for which they are used.
The hook speed for the load hook of the polar crane is 5 feet per minute maximum.
The cask crane's maximum load hook speed is 8 feet per minute.
These speeds are very slow and will not impart major dynamic loads on the rigging.
The cask crane, which is a single-leg gantry running on ore ground level track and one track on the auxiliary building roof is not used to handle heavy loads over the spent fuel area.
The maximum hook speed for the spent fuel bridge crane, which services the spent fuel area, is 7 feet per minute.
This also constitutes a very slow hook speed.
Page 9 FRC Position:
2.1.7 c.
FRC Conclusion The Farley Nuclear Plant complies with Guideline 6.
The crane inspection, testing, and maintenance programs are being revised to satisfy the requirements of ANSI B30.2-1976.
When revision is complete, these programs and appropriate records should be made readily available for review and inspection by the NRC Staff.
Response
The requirements of ANSI B20.2-1976 have been incorporated into FNP procedures.
As with all other plant records, the records associated with crane inspection, testing, and maintenance are 2
available for examination in the Farley Nuclear Plant Document Controi Room by I&E Inspectors.
FRC Position:
2.1.8 c.
FRC Conclusions and Recommendations The Farley Nuclear Plant partially complies with Guideline 7.
To comply with the requirements of this guideline, the Licensee should perform a point-by-point comparison of the 1976 and 1967 revisions of ANSI B30.7 and resolve any issues which do not satisfy the requirements of the 1976 revision.
Response
l A point-by-point comparison of ANSI B30.2-1967 and ANSI B30.2-1976 has been completed with the exception of the demineralizer hoist and the pnlar crane.
The demineralizer monorail hoist is a monorail not a crane and is not covered by either standard.
An
Response (cont'd)
Page 10 evaluation is currently underway to determine it the polar crane satisfies the additional requirements of the 1976 standard.
The results of the study are scheduled to be supplied to the NRC by June 1, 1982.
e FRC Position:
j 2.2.1 FRC Conclusions and Recommendations Since the spent fuel handling crane is lot
- ingle-failure-proof, and the weight allowed by plant technical specifications exceeds +ne maximum heavy load that NUREG-0612 allows to be handling over fuel in the storage pool, the Farley Nuclear Plant does not comply with Interim Protection Measure 1.
To comply, the Technical Specification should be modified to prohibit heavy load movement in excess of the weight of the fuel assembly and gripper tool (2500 pounds).
Response
f The spent fuel pool crane Technical Specifications load limit is i
based on the weight of the spent fuel transfer slot gate which has 1
been discussed in our response to FRI Position 2.1.2.
FRC Position:
2.2.2 b.
FRC Evaluations, Conclusions, and Recommendations FRC's evaluations, conclusions, and recommendations are I
contained in discussions of the respective general guidelines in Sections 2.1.2, 2.1.3, 2.1.4, and 2.1.7.
1 i
A Page 11
Response
Alabama Power Company's responses _ to these itens are included in the appropriate section.
FRC Position:
2.2.3 c.
FRC Conclusion The Farley Nuclear Plant does not comply with the criteria of Interim Protection Measure 6 because the License has not indicated that the special attention and special reviews identified in this interim measure have been addressed for the handling of heavy loads over the core.
The Licensee should take appropriate actions to satisfy the special review ar.d special attention criteria identified.
Response
Alabama Power Company's May 15, 1981 response was not intended as an indication that only routine attention is paid to lif ts over the core.
These lifts are mininized when possible and receive extensive attention when they must be made.
This careful attention j
to detail for the lifts made over the core has been a part of FNP l
operations since the initial plant operation.
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