ML20010H990
| ML20010H990 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 09/22/1981 |
| From: | Clayton F ALABAMA POWER CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0615, RTR-NUREG-615, TASK-A-36, TASK-OR NUDOCS 8109290478 | |
| Download: ML20010H990 (14) | |
Text
Malling Address Alabima Power Company 600 N0rth 16*n Street Post Off ace Box 2641 Birmingham, Alabama J5291 Telephono 205 783-6081
)
F. L. Claytoa, Jr.
Senior Vice President yghggg gig Flintridge Ou:!d:ng liv 5ckJffNYn Vht!!CSffum September 22, 1981 g's Docket No. 50-348
$7 No. 50-364 t
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Director, Nuclear Reactor Regulation A
Qj U. S. Nuclear Regulatory Connission
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Washington, D. C. 20555 O'
s s,
- N Attention
- Mr. D. G. Eisenhut Joseph M. Farley Nuclear Plant - Units 1 and 2 NUREG 0612 Control of Heavy Loads at Huclear Power Plants Gentlemen:
In response to Mr. Eisenhut's letter dated December 22, 1980 concerning the above subject, Alabama Power Company hereby submits the enclosed.
If you have any questions, please advise.
Yours very t ly,
[
/
' llL,
,1
/
Enclosure cc: Mr. R. A. 'homas Q33 Mr. G. "
frowbridge 5
(w/ enclosure)
Mr. 1 O'Reilly(w/ enclosure) c.
Mr. E. A. Reeves Mr. W. H Bradford (w/ enclosure)
/ /
8109290478 810922 PDR ADOCK 050CO348 p
bc: Mr. W. O. Whitt Mr.-R. P. Mcdonald Mr. H. O. Thrash (w/ enclosure)
Mr. O. D. Kingsley, Jr. (w/ enclosure)
Mr. W. G. Hairston, III (w/ enclosure)
Mr. J. W. McGowan (w/ enclosure)
Mr. C. D. Nesbitt (w/ enclosure)
Mr. R. G. Berryhill (w/ enclosure)
Mr. D. E. Mansfield (w/ enclosure)
Mr. J. A. Ripple (w/ enclosure)
Mr.W.C.Carr(w/ enclosure))
Mr. J. R. Crane (w/ enclosure Mr. L. B. Long (w/ enclosure)
Mr. A. A. Vizzi (w/ enclosure) i 3
(
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.. ~,.,__.
Page 1 of 12 RESPONSE 10 STAFF POSITION SPECIFIC REQUl))EMEllfS FOR OVERilEAD llANDLit!G SYSTEllS Note:
Staff Positions are quoted from Enclosure 3 to the December 22, 1980,ilRC letter on Control of Ileavy Loads.
Staff Position:
2.2 SPECIFIC REQUIREMENTS FOR OVERilEAD !!ANDLING SYSTEfG OPERATING IN THE VICINITY OF FUEL STORAGE P0OLS ilVREG-0512, Section 5.1.2, provides guidelines concerning the design and operation of load-handling systems in the vicinity of stored, spent fuel.
Information provided in response to this section shouid demonstrate that adequate neasures 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 will not exceed the limits set by the evaiuation criteria of flVREG-0612, Sec^ ion 5.1, Criteria I threugh III.
1.
Identify by name, type, capacity, and equipment designator any cranes physically capable (i.e., ignoring interlocks, moveable mech-
~~
anical stops, or operating procedures) of carrying loads which could, if dropped, land or fall into the spent fuel pool.
Respong:
The only cranes which can carry loads over the spent fuel pool is the spent fuel pool bridge crane and the cask har.:iling crane.
The only load which the spent fuel pool bd dae erane can handle is a f el assembly which is not a heavy load per the "hhavy load" definition in Chapter 1 of NUREG-0612.
Since thn Unit 1 and 2 oparating licenses prohibit the use of the cask crane to move casks r.o crane is capable of carrying a heavy load over the spent fuel pool and therefore, neither are currcntly applicable to the flVREG-0612 concerns.
Staff Position:
2.
Justify the exclusion of any cranes in this area from the above category by verifying that they are incapable of carrying hcavy loads or are permanently prevented from movement of the hook centerline closer than 15 feet to the pool bourdary, or by providing a suitable analysis demonstrating that for any failure mode, no heavy load can fall into the fuel-storage pool.
Response
e There are no cranes in the vicinity of the spent fuel pool that are capable of transporting a heavy load over the pool.
Page 2 of 12 RESPONSE TO STAFF POSITION SPECIFIL REQUIREMENTS.FOR OVERHEAD h..lDLING SYSTEMS Stjaff Position:
3.
Identify any cranes listed in 2.2-1, above, which you have evaluated as having sufficient design features to r uke the likelihood of a load drop extremely small for all loads to be carried and the basis for this '
evaluation (i.e., complete compliance wi th 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
Not applicable to Farley Nuclear Plant.
Staff Position:
4.
For cranes identified in 2.2-1, above, not categorized accordir.g to 2.2-3, demonstrate thut the criteria of NUREG-0612, Section 5.1, are satisfied.
Compliance with Criterion IV will 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 scent i ~
fuel area and your determination of complianca. This response should include the following infonnation for each crane:
Which alternatives (e.g., 2, 3, or 4) from those identified in l;UREG-a.
0612, Section 5.1.2, have been selected.
Response
^
The cask han:lling crane at Farley Nuclear Plant is not presently in service except as an outside auxiliary crane.
The Unit 1 operating license prohibits this crane from being used to move casks.
_Staf f Positite:
- 2. 3 SPECIFIC REQUIREMENTS OF OVERHEAD HANDLING SYSTEMS OPERATIliG IN THE CONTAINMENT NUREG-0612, Section 5.1.0. provides guidelines concerning the design and operation of load-handling systems in iSe vicinity of the reactor core.
Infonnation provided in response to this section should be sufficient to 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 will not exceed the limits set by the evaluation criteria of MUREG-0512, Section 5.1, Criteria I through III.
RESPONSF TO STAFF POSITION Page 3 of 12 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING _ SYSTEMS 1.
Identify by name, type, capacity, and equipment designator any cranes physically capable (i.e., taking no credit for any interlocks or operating procedures) of carrying heavy loads over the reactor vessel.
Response
The cranes listed below are the only cranes capable of carrying heavy loads over the reactor vessel or the Rod Cluster Control (RCCl change fixture.
Unit Name
- /pe Capacity Equipment Designation 1
Polar Crane Bridge 140 ton Q1T31K001-N 1
Jib 10,000 lb.
N1T31K005 1
Jib 10,000 lb.
N1T31K006 2
Polar Crane Bridge 140 ton Q2T31K001-N 2
Jib 10,000 lb.
N2T31K005 2
Jib 10,000 lb.
N2T31K006 Staff Position:
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 the movement 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 may lanc' in or on the reactor vessel.
R_esponse:
There are no other cranes which are physically capable of lifting a heavy load over the reactor vessel.
Staff Positi,on:
3.
Identify any cranes listed in 2.3-1, above, which you have evaluated as having sufficient design features to make the likelihoo4 of a load drop extremely small for all loads to be carried and the bisis for this evaluation (i.e., complete compliance with NUREG-0612, Se. tion 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.
Responso:
The Polar crane and the Jib cranes are not single-failure-proof, so this section does not apply to Farley Nuclear Plant.
RESPONSE TO STAFF POSITION Page 4 of 12 SPECIFIC REQUIREMENTS _ FOR OVERilEAD llANDLING SYSTEMS Staff Positi_on:
4.
For cranes identified in 2.3-1, above, not categorized according to 2.3-3, demonstrate that the evaluation criteria of NUREG-0612, Section 5.1, are satisfied.
Compliance with Criterion IV will be demonstrated in your response to Section 2.4 ef this request. With respect to Criteria I through III, provide a discussion of you evaluation of crane operation in the containment and your datermination of compliance.
This response should include the following information for each crene:
a.
Where reliance is placed on the installation and use of electrical interlocks or mechanical stops, indicate the circumstances under which these protective devices can be removed or bypassed and the administrative procedures invoked to ensure proper authorization of such action.
Discuss any related or proposed technical specifica-tion concerning the bypassing of such interlocks.
Response
The polar crane is the service crane for the containment and must be capable of servicing the entire refueling floor.
It is not possible, therefore, to restrict its movement via electrical interlocks or mechanical stops.
The Jib cranes have been administratively prevented from lifting heavy loads over the reactor vessel while the vessel head is removed.
Staff Position:
v' b.
Where reliance is placed on other, site-specific considerations (e.g., refueling sequencing), provide present or proposed technical specifications and discuss administrative or physical controls provided to ensure the con +inued validity of such considerations.
Response
The prevention of heavy load drop onto the RCC change fixture is accomplished by administrative controls. The upper internals and the RV head will have been placed in their storage area before the fixture is put in use (i.e.,
contains spent fuel assemblies).
The refueling will be completed before the internals at._ Mad are lifted from their storage area.
Furthermore, the RCP motor, polw crane load block, and up9er internals will be prevented from traveling over the RCC changing fixture during refueling since the fixture is an exclusion area and as such is not part of the safe load path for the polar crane.
The jib cranes have been administrativeiy prevented from lifting heavy loads over the reactor vessel while the vessel head is removed.
The reactor vessel is also protected by administrative controls. When the RV missile shield is in place, the vessel is protected from potential heavy -
loads (Reactor Coolant Pump Motor or crane load block) by the shield.
The reactor vessel head and internals will only be removed from the reactor vessel during refueling.
Prior to removal of the internals, however, the reactor vessel missile shield and reactor vessel head must be removed and placed in their re-spective laydown areas.
When the reactor vessel missile shield is removed, the reactor vessel becomes an exclusion area and is no longer part of the heavy loads safe load path.
Therefore, since no heavy loads will be handled over the reactor vessel during r efueling, a heavy load drop into the reactor vessel is precluded.
RESPONSE TO STAFF POSITION Page 5 of 12 SPECIFIC RI.QUIREMff'TS FOR OVERl!EAD HANDLING SYSTEMS Staff Position:
c.
Analyses perfonned to demonstrate compliance with Criteria I through III should conform with the guidelines 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 appro-priate, for each analysis performed.
Criterion 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 equal to or less than 1/4 of Part 100 limits).
Response
The dropping of a heavy load onto the RCC change fixture or into the reactor core is prevented by the definition of safe load paths. There fore,
no significant release of radioactivity would result from dropping such a load.
Staff Position:
Criterion II. Damage to fuel and fuel storage racks based on calculations involving accidental dropping of a postulated heavy load does not result in a configuration of the fuel such that k rf is larger than 0.95.
e
Response
In the event a 1.ew fuel assecly in the RCC changing fixture is crushed, Keff will remain less than.95 since the refueling canal will be filled vith 2000 ppm barated water during refueling.
For Farley Nuclear Plant cores, it has been determined that for a new core in 2000 ppm borated water with control rods fully inserted, the Keff is.9 or less.
Per Section 4.2.2 "Neutronic Analyses for a PWR Core" of Appendix A of NUREG-0612 the maximum reactivits insertion due to crushing is.05.
There-fore, the maximum achievable Keff would be less than.95.
Staf f Position:
Criterion Ill.
Damage to the reactor vessel or the spent fuel pool based on calculaticos of damage following accidental dropping of a postu-lated heavy load is limited so as not to result in water leakage that could uncover the fuel (makeup water provided to overcome leakage should be from a borated source of adequate concentration if the water being lost is borated).
Response
The dropping of a heavy load either into or onto the reactor vessel is prevented by the definition of safe load path 3.
Page 6 of 12 RESPONSE TO STAFF POSITIC'!
SPECIFIC REQUIREMEllTS FCR OVERHEAD liANDLING SYSTEMS Staf_f Position:
2.4 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN PLANT AREAS CONTAlflING EQUIPMENT REQUIRED FOR REACTOR SilUTDOWN, CORE DECAY HEAT REMOVAL, OR SPEllT FUEL POOL COOLING NUREG-0612, Section 5.1.5, provides guidelines concerning the design ar.d operation of load-handling systems in the vicinity of equipment or coaonents required for safe reactor shutdown 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 which might prevent safe reactor shutdown or prohibit continued decay heat removal is extremely small, or that damage to such equipment from load drops will be limitod 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.
Response
a.
Cortninment Building-In the containment building only the polar crane can carry heavy loads over safety-related equipment.
The safety-related equipment in the contain-ment which could be impacted by a ner vy load drop is the pressurizer, steaa generators, and the reactor coolant pump.
During modes 5 and 6, the pressurizer and steam generators are not required to maintain a safe shutdown condi tion.
Therefore, no load drops onto the pressurizer or steam generators durin9 modes 5 and 6 have been considered.
However, the RCP motor could be dropped back onto the reacto coolant pump while performing maintenance.
For modes 1-4, the polar crane load block is the only heavy load which need be considered since no other heavy loads in containment would be moved during modes 1-4.
b.
Auxiliary Building In the auxiliary building, the demineralizer hatch monorail hoist could drop the demineralizer hatch onto protions of the boric acid transfer system.
The hatch is comprised of two halves, both of which weigh in excess of four (4) tons. This is the only crane / load combination in the auxiliary building which could impact on safety-related equipment. Alabama Pcwer Company's June 24, 1931 NUREG 0612 response provides specific information regarding this crane / load combination.
Staf f Position:
i l
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 i
l extremely small for all loads to be carried and the basis for this evaluation (i.e., complete compliance with NUREG-0612, Section 5.1.5, 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) infonnation specified in Attachment 1.
Response
This section is not applicable to Farley Nuclear Plant.
Pace 7 of 12 RESP 0:lSE TO STAFF POSITIOil SPECIFIC REQUIREMGtTS FOR OVERHEAD !!A4DLl!!C SYSTEMS
_ Staff Positi_on:
E.
For any cranes identified in 2.1-1 not designated as single-failure-proof in 2.4.1, a comprehensive hazard evaluation -hould be provided which includes the following information:
a.
The presentation in a matrix format of all heavy loads and potential impact areas where damage might occur to safety-related equipment.
Heavy loads identification should include designation and ueight or cross-reference to information provided in 2.1.3-C.
Impact areas should be identified by construction zones and elevations or by some other method such that the in. pact area can be located on the plant general arrangement drawings.
Figure 1 provides a typical matrix.
Response
The matrices for all heavy loads and potential impact areas addressed in part 1 above are shown in Figures 1 and 2.
Staff Position:
b.
For each interaction identified, indicate which of the load impact area combinations can be eliminated because of separation and redundancy of safety-related equipment, mechanical stops and/or electrical interlocks, or other site-specific considerations.
Elimination on the basis of the aforementioned considerations should be supplemented by the follcwing specific information:
(1)
For load / target combinations eliminated because of separation and redundancy of safety-related equipment, discuss the basis for de-termining that load drops will not affect continued system operation (i.e., the ability of the system to perform its safety-related func tion).
Response
a.
Containmen t_ Building:
A heavy load drop onto one steam generator would not prevent the plant from coming to a safe shutdown condition since the system could still perform the safety function.
b.
Auxiliary Building:
Dropping the demineralizer hatch onto the boric acid transfer pump and its associated piping could at most disable one train of the system therefore leaving one train to perform the required safety function.
Page 8 of 12 RESPONSE TO STAFF POSITION SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS
_ Staff Position:
(2) Where mechanical stops or electrical interlocks are to be provided, present details showing the areas where crane travel will be prohibited. Additionally, provide a discussion concerning the procedures that are to be used for authorizing the bypassing of interlocks or removable stop, 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 completed.
Respon_se:
No mechanical stops or interlocks are proposed.
Staff Position:
(3) Where load / target combinations are eliminated on the basis of other, site-specific considerations (e.g., maintenance sequenc-inj), provide present and/or proposed technical specifications and discuss administrative procedures or physical constraints invoked to ensure the continued validity of such considerations.
Response
This option was not utilized in the Farley Nuclear Plant.
Staff Position:
For interactions not eliminated by the analysis of 2.4-2-b, above, c.
identify any handling systems for specific loads which you have evaluated as having sufficient design features to make the likelihoed of a load drop extremely small and the basis for this evaluation (i.e.,
complete compliance with NUREG-0612, Section 5.1.6, or partial compliance stpplemented 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.
Pgsyonse:
inis does not apply to Farley Nuclear Plant.
Staff Position:
d.
For interactions not eliminated in 2.4-2-b or 2.4-2-c, above, dem-onstrate using appropriate analysis that damage would not preclude operation of sufficient equipment to allow the system to perform its safety function following a load drop (NUREG-0612, Section 5.1, Criterion IV).
For each analysis so conducted, the following information should be provided:
RESPONSE TO STAFF-FnSITION Page 9 of 12 SPECIFIC REQUIREMENTS FOR OVERHFA.D HANDLING SYSTEMS (1) An indication of whether or not, for the specific load being investigated, the over-head crane-handling system is designed and constructed such that the hoisting system will retain its load in the event of seismic accelerations equivalent to those of a safe shutdown earthquake (SSE).
(2) The basis for any exceptions taken to the analytical guidelines of NUREG-0612, Appendix A.
(.1) The infonnation requested in Attachment 4.
Criterion IV.
Damage to ee,aipment in redundant or dual safe shutdown paths, based on calculatio,is assuming the accidental droppino of a postulated heavy load, will be limited so as not to result in loss of required safe shutdown functions.
Re_spon se:
A missile shield grating covers the pressurizer cavity to protect other equipment from possible pressurizer missiles.
It has been shown by analysis however that this missile shield will also withstand a free fall drop of the polar crane load block. Therefore the pressurizer is protected frcm a heavy load drop during modes 1-4.
It has also been determined that a drop of the reactor coolant pump motor back onto the reactor coolant pump will not compromise the integrity of the RCS loop pressure boundary.
)
These are the only heavy loads not climinated in 2.4-2-b above.
W
i I
Nge 10 of 12 L0 ADS Polar Crane Load Block
, Domineralizer Hatch RCP Motor 9.003 lbs.
(9,425 lbs. + 8,S25 lbs. = 18,253 lbs.)
77,300 lbs.
Related l Elimination Sa fe ty-Hazard Safety-Hazard Safety-Hazard Location Impact Area i
Related Elimination Related Elimination Elevation Ecuiomenti Category Elevation Eautement Category Elevation Ecuipment Category Centainment 34' x 295 CW from 170' 4" Pre >>ar-e j
00 Azimuth izer l
CV feca 190' 0" l Steam b
I
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32' g 25 Contairment O Azimuth
~
! Generator jA I
i l
I l
Coctairment 32' x 265 CW from 190'0"l Steam l
b 6
0 Azimuth Generator l
0 B
I 4
i i
0 Centainment 32' x 145 CW from 193' 0" Steam b
0 0 Azimuth Generatori C
[
i l
i Containment 39' x 120 CW From 155' 0" Reacter !
e 00 Azimuth j
\\
/
Ccolant l
l Pump i
l Containment 39' x 235 CW From l
}
155' 0" Reactor e
0 Azimuth Coolant 0
4 Containment 35' x 355 CW From 155' 0" Reactor e
0 Azimuth Coolant 0
Pump Auxiliary 2' West of Inter-100' 0" Coric b
Building section of Column Acid Lines N and 12 Transfer Pwp
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FIGURE 1 - U*i!T 1 LOAD / IMPACT AREA MATRIX Notes for Figures 1 and 2 are on Page 12 of this Attachment.
I Pap 11 of 12 L0 ADS l
Folar Crme Lead Bleck i
Cemi " ralizer Hat:h I
FC' Mater 7.000 los.
i (9,423 lbs, + P 325 7bs. = 13,253 lbs.)
I 77.20 lbs.
S3fety-l H :ard l
Sa fe ty-I hazard i S3fety. } Hazard l Pelated i
l limin rien Related
' Related Eli
.3ti lElinination E!cntion !Equipr ent ( Category Location l
I pact Area l Elevatie:c Cuier'unt j Category Elevation E c.uirm" t Cate: cry 1
f 8
j 170' 4" i Pressur-l l
'l l
I Containment 34' x 115 CW e
j l
1:er
)
i i
i l
i I
i I
I i
i i
i i
I i
Sntainment l
32' x 205 CW j
19T 0"j Steam b
I I
0
.Cecerateri i
i l
l iA i
i I
'l i
a 1
I I
i l
+
l i
U Containment 32' x 85 CW 190' 0" Steam b
I l
i j Ger.erato-i f
l iS l
6 i
1 i
i l
l l
32' x 325 CW 193'0"l Steam l
b
{
Containment l Generater' I
I f
'C i
1
}
I i
t n
i i
Contahnent 3T t 120" CW From l
i i
155' 0"
'Reacter e
00 Azimuth I
! Ccolant
{
! Pu no l
l l
l Certainment 39'3 235 rW From I
l 1E5' 0" lCc3ctor e
0 Azimut:1 l
l
!C clant l l
l
.F I
l um?
l Centsinmer.t 35' x 355 CW From I
155' C"
! Reactor I e
00 Azimuth l
jCrclant I l
Pt=p i
I i
Auxiliary 2' West of b.ter-l 13T 0" r.oric b
Cuilding
- cction of Column i
Acid Line: N and PG
'r m s fer i
'u p I
l l
i___._.._...._
1' 1
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f1CTE ? -- tiNIT ? 1.0AD/ IMPACT TSCA MATRIX Notes for fives 1 and 2 are en Page 12 of this Attachment.
t
Page 12 of 12 fictes for Figure. 1 and 2:
a.
Crane travel for this area / load combination prohibited by electrical interlocks or mechanical stops.
b.
System redundancy and separation precludes loss of capability of system to perform its safety-related function following this load drop in this area.
c.
Site-specific considerations eliminate the need to consider load /
equipment combination.
d.
Likelihood of handling system failure for this load is extremely small (i.e., Section 5.1.6, NilREG-0612, satisfied).
Analysis demonstrates that crane failure and load drop v;ill not damage e.
safety-related equipment.
.