ML19290C301
| ML19290C301 | |
| Person / Time | |
|---|---|
| Issue date: | 08/27/1979 |
| From: | NRC - NRC THREE MILE ISLAND TASK FORCE |
| To: | |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, TASK-A-36, TASK-OR NUDOCS 8001100505 | |
| Download: ML19290C301 (30) | |
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TABLE 1 SURVEY OF HEAVY LOADS OVER (0) OR ONLY PR0XIMITY TO APPR0X.
FREQUENCY AREA LOADS HANDLING SPENT FUEL WEIGHT (tons)1/
HANDLED
- 1. PWR - Refueling 1.
Spent Fuel Shipping Cask (P)15-110 Tons 2]
Building 2.
Pool Divider Gates (some plants)
(P) 2 Tons 2-4 x's (per refueling) 3.
Fuel Transfer Cana.1 Door (P) 2 Tons 2-4 x's (per refueling) 4.
Missile Shields (P) 4-20 Tons 2 x's (per refueling) 5.
Irradiated Specimen Shipping Cask (P) 3.5-12 Tons Once per year to once per 10 years 6.
Plant Equipment (some plants)
(0) 2-4 Tons As required for modification or replacement 7.
Spent resin, filter, or other (P) 5-10 Tons
- 5 x's per year radioactive material shipping casks 8.
New fuel shipping containers with fuel (usually 4 assemblies)
(P) 3 Tons 3/
9.
Failed Fuel Container (0) 1 Ton less than once per refueling
]
- 10. Fuel transfer carriage
(:0) or (P) 1.5 Tons Only for main-tenance or repair CD
( once per 10 year
- 11. Crane Load Block (0) 4-10 Tons 5/
w 2.
PWR - Containment 1.
Reactor Vessel Head (0) 55-165 Tons 2 x's (per refueling)
Building 2.
Upper Internals (0) 25-65 Tons 2 x's (per re-fueling) 3.
In-Service Inspection Tool (0) 4.5 Tons Used at least once every three years
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FREQUENCY AREA LOADS HANDLED SPENT FUEL WEIGHT (Tons)1/
HANDLED 3.(cont.)
- 12. Waste and Debris Shipping (0) (over 8-30 Tons 1-3 x's (per year)
Casks reactor and/
or spent fuel pool)
- 13. Vess~el Head Insulation (P)
A-6 Tons 2 x's (per refueling)
- 14. Replacement Fuel Storage (0) (over 8 Tons On installation Racks for Spent Fuel spent fuel)
- 15. Crane Load Block (0) 4-10 Tons 5]
4.
Other Plant Areas 1.
Spent Fuel Shipping Casks (0) (over 15-110 Tons 4/
(some plants) safety equip-ment) 2.
Turbine or other equipment (0) (over 2-For equipment in turbine building (sone safety equip-overhaul and plants) ment) replacement
')
N CD W
N LT1
TABLE 1 FOOTNOTES 1/ Listed weight does not include weight of load block which may add 4-10 tons to the weight of the dropped load.
2/ These are presently not being used at most plants.
However, once long term waste repositories are established, casks will be used frequently for shipping spent fuel offiste.
For a typical 1,000 MWe pressurized water reactor, spent fuel casks must be shipped offsite from 7 to 65 tires per year depending on the size c'ask currently licensed for use in the United States.
3/ A typical 1,000 MWe power plant would usually require 16 or 17 new fuel containers (four fuel assemblies each) per year.
4/ These are presently not being used at most plants.
However, once long term repositories are established, casks will be used frequently for shipping fuel offsite.
For a typical 1,000 MJe boiling water reactor, spent fuel casks must be shipped offsite from 12 to 125 times per year depending on the size cask used.
This is based on casks currently licensed for use in the United States.
5/ Since crane load blocks for the larger cranes typically weigh 4-10 Tons.
Thus the load blocks should be considered a heavy load even if they are not carrying a load.
They may be used many times with lighter loads.
I728 326
TABLE 7
SUMMARY
OF LOAD DROP ACCIDENT ANALYSES l
l l
Low Population ~ Zone Minimum No. of Assy's No. of Days Exclusion Radius Dose Subcritical Thyroid Whole Body Thyroid Whole Body ReacQ{o{ Part 100 4 (no. filters:
172.5
.612 17.26
.061 1 Assy 4 (w/ filters) 8.628
.679
.863
.058 12 Assy's P
54 (no filters) 2.319
.0025
.2319
.0002 44 Assy's W
54 (w/ filters)
.1143
.0021
.0114
.0002 8.8x102 Assy's R
90 (no filters)
.1046
.00121
.0105
.0001 9.6x102 Assy's 90 (w/ filters)
.0052
.00120
.0005
.0001 7.0x103 Assy's 120 (no filters)
.0079
.00118
.0008
.0001 7.1 x103 Assy's 120 (w/ filters)
.0004
.00118
.0000
.0001 7.1 x103 Assy's 1 (no filters -
SBGT) 92.260
.701 9.226
.070 1 Assy B
1 (w/ filters-SBGT) 4.613
.653
.461
.065 13 Assy's W
90 (no filters -
R SBGT)
.0360
.0004
.0036
.0000 2.8x103 Assy's 90 (w/ filters -
4 SBGT)
.0018
.0004
.0002
.0000 2.1x10 Assy's I O_ose per fuel assembly damaged (REM's).
2Number assemblies that must be damaged to approach (1/3 of) Part 100 exposure limits of 300 REM Thyroid and 25 REM Whole Body.
/
1728 327 E
To Reach 10 CFR 100 Limits - Spent Fuel Pool Area (PWR's)
.105 104
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Conservative Limit On No. of 3
Assemblies That 10 Could be Damaged l
Realistic Max.
No. of Assemblies That Could Be (220)
Damaged ~ '-
PWR - With Credit i
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for Charcoal I -~ '
2 Fil ters 10
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PWR - No Credit for Char-coal Filters 1728 328 20---
40 - -- 60-- - 80 99 100 - -
120 1C DAYS AFTER SHUTDOWN
TABLE 2 CRANES SATISFYING INTENT OF R.G. 1.104 or NUREG 0554 SINGLE FAILURE PROOF CRANE
{
I YES N0 cx3
- of
- of CNJ
- Plants Cranes
- Plants Cranes r'-
PWR: Containment Polar Crane 0
0 36 36 2
2 Refueling Building Crane 2
2 34 24 BWR: Reactor Building Crane 13 10 53 53 I 4 reactors were covered in the survey.
Not included were eleven (11) SEP plants, two (2) plants indefinitely out of service, two (2) plants which were recently licensed, and Ft. St. Vrain which is a Gas-Cooled Reactor (there are 70 plants licensed to operate as of July 1979).
2However, licensees of three (3) of these plants have committed to upgrade the cranes used to handle the cask (affects two (2) cranes), although no date for completion of this upgrading has been established.
However, one BWR licensee has committed to upgrade the reactor building crane.
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/
TABLE 3
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/I, SURVEY OF DESIGN FEATURES
&o I
RELATED TO CONTROL I
/
0F HEAVY LOADS e
Plants Yes No Containment (Spent Fuel Pool Area) 51 3
Charcoal Filters (Spent Fuel Pool Area) 51 3
Pool Designed for Cask Drop 26 28 Containment Isolation on High Radiation (PWR) 162 20 Compliance with Regulatory Guide 1.13 39 15 I Exclusive of crane features covered by Table 2.
2D.C. Cook 1 & 2 Prairie Island 1 & 2 Ft. Calhoun Rancho Seco Farley Salem 1 Maine Yankee Turkey Point 3 & 4 North Anna 1 Zion 1 & 2 Point Beach 1 & 2 1728 330
TABLE 4
_ SURVEY OF LICENSEE
_ ANALYSES RELATED TO CONTROL OF HEAVY LOADS PLANTS Yes No Cask drop damage to fuel 5
49 Fuel handling accident 54 0
I Potential for drop to cause criticality 3
51 Plenum assembly or reactor head drop 6
48 I These only considered potential for a drop to cause criticality in the spent fuel pool, but not in the reactor.
1728 33l 9
TABLE 5 SURVEY OF PROCEDURES IN EFFECT RELATED TO CONTROL OF HEAVY LOADS Plants Yesl No2 1.
Procedures on crane operation 34 21 2.
Refueling Procedures 30 25 3.
Movement of Reactor Components during or prior to refueling 27 28 4.
Cask Handling Operations 22 33 5.
Crane operator training 3
52 I In some cases procedures were not submitted, but were referenced by title and/or description.
2Information provided by licensees did not indicate that such procedures were in use.
1728 332
TECHNICAL SPECIFICATIONS g4) T')
gr1 h4A 1.
Plants that do not have a Technical Specification prohibiting handling of heavy loads (i.e., greater than a fuel assembly plus handling tool) or the Cask Over Spent Fuel:
NOTES Big Rock Point Browns Ferry 1 - 3 Single Failure Proof Crane Cooper Limit Switches To Prevent Travel Over Spent Fuel Dresden 1 Duane Arnold Ft. Calhoun Electric interlocks To Prevent Travel Over Spent Fuel Lacrosse H. B. Robinson Single Failure Proof Crane FitzPatrick Monticello Single Failure Proof Crane Nine Mile Point Single Failure Proof Crane Oyster Creek Palisades Pilgrim Maine Yankee 2.
Plants That Prohibit Cask, but do not prohibit smaller heavy loads, from being brought over spent fuel:
Dresden 2 and 3 Single Failure Proof Crane Hatch 1 Single Failure Proof Crane Haddam Neck Indian Point 2 Millstone 1 Quad Cities 1 and 2 Single Feilure Proof Crane Turkey Point 3 and 4 Vermont Yankee Single Failure Proof Crane 1728 333
APPENDIX A-1 g 97 f c[f d
f
,g CAUSES OF CRANE ACCIDENTS DEPARTMENT OF THE NAVY
, ff,0 O
p0 NO. OF LOAD TOTAL NUMBER OF CAUSE DROP OR POTENTIAL CRANE EVENTS RESULTING CATEGORY LOAD DROP EVENTS
% OF TOTAL IN EQUIPMENT DAMAGE
% OF TOTAL 1.
Crane Fail-ure 10 23%
17 23%
(Design Error)
(1)
(2.3%)
(2)
(3%)
(Maintenance Personnel)
(2).
(4.6%)
(2)
(3%)
(Crane Com-ponent Failure)
(7)
(16.3%)
(13)
(17%)
2.
Crane Operator Error 301/
70%
54 73%
( Dis traction /
Inattention)
(11)
(26%)
(24)
(32%)
(Inadequate Training)
(8)
(18%)
(13)
(19%)
(Failed To Follow Pro-per Precautions /
Procedures,)
(11)
(26%)
(17)
(23%)
3.
Rigging 3
7%
3 4%
(Rigger)
(3)
(7%)
(3)
(4%)
(Rigging)
(0)
(0)
(0)
(0) 1/15 (50%) of these events occurred when the crane or hoist was left in the
" raise" mode or inadvertently raised to limit.
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o 17 8 334
APPENDIX A-2 pATA ON CRANE ACCIDENTS I
FROM OSHA RECORDS The following is a statistical sumary of major crane accident causes based on an analysis of over 1,000 crane accidents involving damage to
. equipment:
CAUSE CATEGORY PERCENTAGE A.
Loss of load due to poor rigging or slings 34%
B.
Perfonning minor maintenance, inspection, or unrelated work while load is in motion 22%
C.
Operating crane without authorization or proper signals 18%
D.
Failure of defective boom, cable, or sheaves 14%
E.
Failure due to overloading 4%
F.
Handling load too near stationary equipment 2%
G.
Other causes (including failures of control systems and inadequate inspection or maintenance) 6%
I 0ccupational Safety and Health Administration 1728 335
TABLE 8 r
SURVEY OF CRANE
[
[
LER EVENTS rd
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~~
Cause Category No. of Events A.
Failure during plant. construction phase 2
B.
Failure due to design or fabrica-tion errors 9
C.
Failure due to lack of adequate inspection 2
D.
Failure due to operator error or lack of training 8
E.
Failure due to random mechanical componert failures 5
F.
Failure due to random failures of control cystem components 3
G.
Events due to lack of operating procedures 4
H.
Events due to crane overloading (including load hangup) 1 1728 336
j op % p-u A }k< Hank George / DRAFT 7/26/
,u-e-(I S.Wf h - /C
-7 4.s Adest W M j
N 4.1 Guidelines for Control of Heavy Loads The following sections describe various alternative approaches v:hich provide acceptable measures for the control of heavy loads. The objec-tives of these guidelines are to assure thatIr each areas addressed [~
the-foi l owi n<y-eri t e ria.-a r&s a t-i s f i ed :
(1) Releases of radioactive material that may result from danage
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c>l to spent fuel due to accidental dropping of a h avy load ae.
Ef44 N M k & %
ds.4s4 M@ignificantly less than 10 CFR Part 100 limitsE 7
M 41%de t -
(2) Damage to fuel and fuel storage racks due to accidental dropping of a heavy load will not result in such a configuration of the fuelf that k(eff) approaches or is larger than 1; (3) Damage to the reactor vessel or the spent fuel pool due to s
Q accidentsl dropping of a heavy load will not result in leakage that could uncover the fuel, (makeup provided to overcome leakage should be from a borated source of adequate concentration if the
,\\
water that is being lost is borated)ess;&;
l J
M (4) Accidental load drops will not damage equipment required for safe shutdown or continued decay heat removal such that this equipment cannot perform its intended function.
After reviewing the historical data available on crane operations and identifying the principal causes of load drops, and considering the type and frequency of load handling operations at nuclear power plants, the task group has developed an overall philosophy that provides a defense-hfi-depth approach to controlling the handling of heavy loads.
This philosophy encompasses an intent to prevent as v: ell as mitigate 1728 337
the consequences of postulated accidental load drops.
The following summari$es this defense-in-depth approach:
(1) Provide sufficient operator training, handling system design, load handling instructions, and equipment inspection to assure reliable operation of the handling system; (2) Define safe load paths through procedures and operator training so that heavy loads are not carried over or near spent fuel or safe shutdown equipment; and (3) Provide electrical interlocks to prevent movement of heavy loads over spent fuel or in proximity to equipment fg associ-ated with redundant shutdown paths.
Certain alternative measures may be taken t, compensate for deficiencies in (2) and (3) above, such as the inability to prevent a particular heavy load from being broughtover spent ds.
D#
fuel (e.g., reactor vessel head). These alternative measures (du.a.}N g
can include: increasing crane reliability by providing redundancy in cet=tain components, increased safety factors, and increased inspection as discussed in Section 4.1.6 of this report; or prohibiting crane operations in the spent fuel pool area (PWRs) until fuel has decayed so that off-site releases would be sufficiently low if fuel were damaged.
Even if one of these v e-alternative measures is selected, (1) abogt should still be satisfied, as well as defining safe load paths, to provide adequate defense-in-depth.
1728 538
The following sections provide guidelines on how the above defense-in-depth approach may be satisfied for various plant areas.
Fault trees were developed and used to evaluate the adequacy of the@uidelines and to assure a consistent level of protection for the various areas.
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Hank George / DRAFT 7/26 4.1.1 General All plants have overhead handling systems that are used to handle heavy loadsf in the area of spent fuel in the reactor vessel or in the spent fuel pool.
Additionally, loads may z#o be handled in areas where their accidental drop,may damage safe shutdown systems. Accordingly, skcW S duly h cHegqq all plants for handling heavy 1. oads in the spent fuel pool area, in e.-
containment (PWRs), in,the reactor building (BWRs), and in other(areas where heavy loads NNoughtin proximity %r over safe shutdown systems.
(1) Safe load paths should be defined for the movement of heavy loads to provide maximum practical separation of heavy loads from spent fuel in the reactor vessel and in the spent fuel pool, and from safe shutdown equipment.
These load paths should be defined in procedures or shown on equipment layout drawings.
Deviations from defined load paths should receive prior approval from the shift supervisor or other designated plant management.
(2) Procedures should oe developed to cover critical load handling operations, particularly those requiring use of special handling devices,such as:-removal and replacement of the drywell head (BWRs); removal and replacement of the reactor vessel head; removal and replacement of upper vessel internals; removal and replace-ment of the spent fuel pool gate M, crane operation; and spent fuel cask handling. These procedures should include:
identification of requirea equipment; inspections required before movement of load, including acceptance criteria; critical steps to be followed and proper sequence in handling the load; reference or inclusion of safe load path; and special precautions.
1728 340
. (3) Crane opdr'ators are trained and qualified in accordance with chapter 2-3ofANSIB30.2-1976,"0vt-headandGantfCranes,"
prior to their handling of heavy loads.
(4) Special lifting devices should meet the intent of ANSI N14.6-1978,
" Standard for Spe'cial Lifting Devices for Shipping Containers Weighing 10,000 pounds (4500 kg) or More for Nuclear Materijh" This standard should apply to all special lifting devices which carry heavy loads in areas as defined above.
For operating plants certain inspections and load tests may be accepted in lieu of certain material requirements in the standard.
In addition, the stress design factor stated in Section 3.2.1.1 of ANSI N14.6 should be based on the combined maximum static and dynamic loads that could be imparted on the handling device This based on characteristics of the crane which will be used.
is in lieu of the guideline in Section 3.2.1.1 of ANSI N14.6 which bases the stress design factor on only the weight (static
}&
load) of the load the intervening components of the special s
handling device, (5) Lif' ting devices that are not specially designed should be installed and used in accordance with the guidelines of ANSI f
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(6) The crane should be inspected, tested, and maintained in
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accordance with Chapter 2-2 of ANSI B30.2-1976, " Overhead and
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e-Gantry Cranes," with the exception that tests and inspections
'f, should be performed prior to ura where it is not practical to meet the frequencies of ANSI B30.2 for periodic inspection and
. : :wu+ -
..c, test, or where frequency of crane use is less than the specified inspection and test frequency. (e.g., the polar crane inside 1728 341
a PWR conta,inment may only be used yearly, or every 18 months, during refueling operations, and is generally not accessible during power operation.
ANSI B30.2, however, calls for certain inspections to be performed daily or crm: Q. p we d y,
For a polar crane these inspections should be performed prior to refueling operations).
(7) The crane should be designed to meet the applicable criteria of CMAA-70, " Specifications for Electric Overhead Cranes."
4.t Alternativey to a specification in CMAA-70 may be accepted in lieu of specific compliance if the intent of the specifi-cation is satisfied.
1728 342
Hank / DRAFT 7/26/79 4.1. 2 SPENT FUEL P0OL AREA - PWR Most PWR's require that the spent fuel shipping cask be placed in the spent fuel pool for loading.
Additionally, other heavy loads may be carried over or near the spent fuel pool, including plant equipment,
,.. v rad-waste shipping casks, damaged, fuel,cM, and the crane load block.
c.4-:m..
To provide assurance that the siuei+ 0f Section 4.1 are met for load handling operations in the spent fuel pool area,4one of the following _
M.3 d.i. A c e -l-ctlah.su e
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should be satisfied:
, je /;.a; e
(1) The crane and associated lifting devices used for handling heavy loads in the spent fuel pool area should satisfy the guidelines of Section 4.1.6 of this report.
OR (2)
Each of the following are provided:
.his alternative 4s httend# fnr nlants_that_do -not-requi re-pl acement -o f--the rasi._
in-4.he-spent fuel pooMor-hadingf:
(a)
Electrical interlocks should be provided that prevent move-
/.s-ment of the crane load block over or within (& feet).i (horizontal) of the spent fuel pool. These electrical interiocks may not be bypassed when the pool contains " hot" spent fuel.ddtAL 4dvdh h[O.bi thi
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(#JE)To preclude rolling if d'ropped,*the cask should be carried at no more than six (6) inches above the floor of the spent fuel pool area.
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Each of the following are provided:
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c 1/28 343 (a) " Hot" spent fuel should be concentrated in one location in the spent fuel pool that is separated as much as possible from load paths.
(b) Electrical interlocks should be provided to prevent movement of the crane load block over or within (25 feet)
(horizontal) of the " hot" spent fuel.
As much as possible loads sHould still be moved over load paths that avoid the spent fuel pool and not brought even as close as,
Inter-(25 feet) to the " hot" spent fuel unless necessary.
locks should not be bypassed as long as the pool contains y-ji
" hot" spent fuel.
These electrical interlocks should not Pec c,
(o,?l be bypassed without approval from the shift supervist::m
.3
,,.r' }r v mf
^ '; a -
(or other designated plant management individual).
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Interlocks should be verified to be operational prior to
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a-h placing " hot" spent fuel in the pool.
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R (d) To preclude rolling, if dropped, the cask should be
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carried at no more than six (6) inches above the floor
/
of the spent fuel pool area.
gd{47 Load drops into the cool # - e m nt i: alle:cd by cle *d:aLint9% should be analyzed and shown not to cause leakage that could uncover the fuel.
OR (4)
Each of the following are provided:
(a) Movement of the crane load block is prohibited if the spent fuel pool contains " hot" spent fuel.
power to the crane is removed during this period.
1728 344 (b) Load drops of heavy loads, including the crane load block, into the pool should be analyzed and shown not to fause leakage that could uncover the fuel.
(c) To preclude rolling, if dropped, the cask should be carried at no more than six (6) inches above the floor of the spent fuel pool area.
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4 OR (5)
The effects of drops of critical loads should be analyzed
- p. f.' id. n and shown to satisfy the critm of Section 4.1 of this report.
Such analyses should c'onsider the following:
i That the load is dropped in an orientation that causes the most severe consequences; ii That fuel impacted is 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> subcritical (or whatever the minimum that is allowed in facility technical specifications prior to fuel handling);
iii. That the load may be dropped at any location in the crane travel area where movement is not restricted by electrical interlocks ;
iv.
That credit may not be taken for spent fuel pool area filters if hatches, wall, or roof sections are removed for handling of the heavy load being analyzedc _-L g,
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Hank / DRAFT 7/26/79 4.1. 3 -
CONTAINMENT BUILDING - PWR
- P,.R containment buildings contain a polar crane that is used for removing and reinstalling shield plugs, the reactor vessel head, upper w s >.>.efe)%-
-W intervals, andgeactor coolant pump fi-f%W ).
Other heavy loads that the polar crane may be required to handle are the reactor vessel
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inspection platform and the damaged fuel cask.
To provide assurance (that the criteria of Section 4.1 are met for load handling operations
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H e e 'se
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,.J !, g'll.in the containment building,Yone of the following should be satisfied:
'W a.c i c,, u !. / ) (1 )
The crane and associated lifting devices used for handling heavy L
loads in the containment building should satisfy the guidelines of Section 4.1.6 of this report.
OR (2)
The effects of drops of heavy loads should be analyzed and AWa-u -
shown to satisfy the crutra of Section 4.1.
Loads analyzed should include the following: reactor vessel head; upper vessel internals; vessel inspection platform; damaged fuel cask; irradiated sample cask; reactor coolant pump (unless safe load paths are defined and electrical interlocks are provided to prevent movement of the load block over or near the reactor rc vessel when handling a reactor coolant pump en other plant equipment); any other heavy loads brought over or near the reactor vessel.
Loads need not be analyzed if their ef fects are scoped by the analysis of some other load.
In this analysis, credit may be taken for containment isolation sf N pu g
- = i f such is provided with prompt automatic actuation on high radiation.
Additionally, the analysis should satisfy 4.1.2(5) i-iii.
1728 346
Hank / DRAFT 7/30/79 i
4./.4 Reactor Building-BWR The reactor building in BWRs typically contains the reactor vessel and spent fuel pool, as well as various safety-related equipment.
i.n The reactor building overhead crane may be used many day-to[ day g
operations such as moving waste shipment casks or handling plant equipment related "to maintenance or modification activities.
The crane may also be used during refueling operations for removal and reinstallation of shield plugs, drywell head, reactor vessel head, P%' '-rd steam dryers and separat>y,ons, and refueling canal gates.
The crane would also be used subsequent to refueling for handling of the spent cu k fuel shipping cask.
This eeble may be lifted as high as feet above the floor elevation at which the cask is brought into the reactor building.
The assure that the criteria of Section 4.1 are satisfied, one of the following should be met:
(1)
The reactor building crane, and associated lifting devices used for handling the above heavy loads, should satisfy the criteria of Section 4.1.6 of this report.
OR (2),The effects of heavy load dgrops in the reactor building should be analyzed to show that the h,e c% es eV is of Section 4.1 are satisfied.
The loads analyzed should include:
shield plugs, drywell head, reactor vessel head; steam dryers and separators, refueling canal (bp gates, spent fuel shipping cask; vessel inspection platform, and any other heavy loads that may be brought over or near spent fuel in the reactor vessel or spent fuel pool.
A Credit may be taken in this analysis for operation of the standby Gas Tre atment $ystem if facility technical specifications require its 1728 347
_2 Icperation during periods when the load being analyzed would be handled.
The analysis should also conform to items 4.1.2 (5)[through iii.
Loads whose effects are scoped by the analysis of some other load need not e analyzed.
4.1.5 Other Areas In other plant / areas, loads may be handled which, if dropped in a certain location, may damage safe shutdown equipment.
Although this is not a concern at all plants, loads that may damage safe shut-down equipment at some plants include the spent fuel shipping cask,
$binegeneratorpartsintheturbinebuilding,andplantequipment such as pumps, motors, valves, heat exchangers, and switchgear.
Some of these loads may be smaller than a " heavy load", but may be sufficient to damage safe shutdown equipment.
'In addition, the cask may be handled at heights greater than the 30 feet
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at which the cask is qualified for a load drop, such as in the trans-
/
porting the cask from the spent fuel area' to the vehicle which will N
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transport the cask offsite. A drop of the cask may result in damage I
ktothecaskandN1 gas of its contents.
.e For plant areas other than reactor and containment buildings and spent fuel pool areas, the following should be satisfied for cranes that may carry loads over safe shutdown equipment or that may handle the spent fuel shipping cask:
(1)
If the spent fuel cask can be handled at a height of 30 feet or
~
more above the ground (or above a floor), then one of the following should be met:
1728 348
. m t
^
The crane and associatedpifting device should conform to a.
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the criteria of,Section 4.1.6 of this report, 0%
b'.
The effects of a drop of the cask fron its maximum height -
s
'should be. analyzed and shdw that the releases wou.ld be
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significantly less than 10 CFR Part 100 limits. -
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If safe shutdown equipment are within the trayel path of overhead )@
( e r-a.~.y ch v A ;d h d c'q c7 $ c.e J.;wc.,.s handling systems (suchJ carrying the cask over a safety related building or handling of a main turbine rotor over a safety related area in the turbine building) one of the following should be satis-fied:
The crane and associated lifting device should conform a.
to the criteria of Section 4.1.6 of this report; O R, b.
Safe load paths should be defined so that loads are not carried in proximity to safe shutdown equipment or cabling such that a load drop could impact this equipment.
If the load could impact equipment or cabling associated with redundant safe shutdown paths, electrical interloc should be provided to prevent movement of loads in proximity to these redundant safeh Y
oR The effects of load dropshave been analyzed and are such that c.
damage to safe shutdown equipment would not preclude operation of the equipment to perform its intended function.
1728 349
4.1.6 Single Failure Proof Handling Systems For certain areas, to meet the guidelines of Sections 4.1.2, 4.1.3
~
4.1.4, or 4.1.5, the alternative of upgrading the crane and lifting devices may be chosen. The purpose of the upgrading is to improve the reliability of the handling system through increased safety factors m e, t
and reddn,dancy in certain active components.
NUREG - 0554, " Single A
Failure Proof Cranes For Nuclear Power Plants" provides guidarce ktrw Aiew for design, installation, and testing of new cranes that are of a high reliability design.
For operating plants,. Appendix C to this report provides guidelines on implementation of NUREG - 0554 for operating plants and plants under construction in lieu of satisfying all-criteria of NUREG - 0554.
There are no staff guidelines issued on design or use of special handling devices or slings. Section 4.1.1 of this report provides certain guidance on special handling devices and slings. Where the alternative of upgrading the sling is chosen, then steps beyond section 4.1.1 should be taken.
delws
,p TLer<bt,T'he following should be met where the alternative te upgradbandling system reliability is chosen:
(1)
Lifting Devicei:
(a) Special Lifting Devices that are used for heavy loads in the e
area where crane is to be upgraded should meet ANSI N14.6 s
1978, " Standard For Special Lifting Devices for Shipping Containers Weighting 10,000 pounds (4500 kg) or More For Nuclear Materials" as specified in Section 4.1.4(4) J (u. n p -
exceptthattheloade..,Mnis,, defined as a " critical" load.
1728 350
i (b)
Lifting Devices that are not specially designed and that are used for heavy loads in the area where the crane is to be upgraded should meet ANSI B30.9 - 1971, " slings" as specified c= un e.a.
in Section 4.1.4(5)3except hat one of the following shou'd also be satisfied:
(i)
Provide redundant slings such that a single component
._y failure or malfur.ction in the sling will not result in uncontrolled lowering of the load; OR gL c, A4 (ii)
Load-be.aring members of the sling Qbe provided with stress design factors that are twice what is called for in
_ _s, meeting 4.1.4(5) of this report.
2 New Cranes should be designed to meet NUREG - 0554, " Single
- a.3,\\.
Failure Proof Cranes For Nuclear Power Plants." For operating
'/
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plants or plants under construction, the crane should be
- yr
/,-
upgraded in accordance with the implementation guidelines of r.kc/
/
Appendix C of this report.
./
r' 3.
Interfacing lift points such as lifting lugs or cask f ruwien s should also meet one of the following:
(a) Provide redundancy such that a single lift point failure will not result in uncontrolled lowering of the load; lift points should have a stress design factor of five (5) times the maximum combined static and dynamic load.
OR p
(b) L4ft pointi[fiave a stress design factor of ten (10) times the maximum combined static and dynamic load.
e 1728 351