ML20040C232
| ML20040C232 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 01/22/1982 |
| From: | Withers B PORTLAND GENERAL ELECTRIC 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 TAC-07716, TAC-7716, NUDOCS 8201270436 | |
| Download: ML20040C232 (18) | |
Text
-
9 4
=
la E ortland General BectricConyxurf 8EcaiyED
\\'
P g,
i
'{# g= ane imb ~p J
artDwem W.e ne91ert 0
on V
January 22, 1982 Trojan Nuclear Plant Docket 50-344 License NPF-1 Mr. Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20555
Dear tir. Eisenhut:
Control of Heavy Loads, NUREG-0612 Attached for your review are the itemized responses to Sections 2.2, 2.3 and 2.4 of Enclocure 3 of your letter of December 22, 1980 which forwarded NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants".
This letter forwards results of evaluations and analyses of the load handling practices at the Trojan Nuclear Plant to demonstrate that the risk of handling heavy loads is extremely small.
Sincerely,
=_ e Bart D. Withers Vice Presidert Nuclear Attachments c:
Mr. Lynn Frank, Director 33 State of Oregon y
Department of Energy II 8201270436 820122 PDR ADOCK 05000344 P
PDR 121 sW Sirmo Sweet Rytra Ovxn 97204
Troj:n Nucionr Picnt Attachmsnt 1 Dockst 50-344 Page 1 of 3 License NPF-1 PGE'S RE WONSE TO INFORMATION REQUESTED BY THE NRC STAFF IN SECTION 2.2 0F ENCLOSURE 3 TO NRC LETTER OF DECEMBER 22, 1980 1.
Identify by name, type, capacity, and equipment designator any cranes physically capable (ie, ignoring interlocks, moveable mechani-cal stops, or operating procedures) of carrying loads which could, if dropped, land or fall into the spent fuel pool.
PCE Response:
Equipment Name Type Capacity Designator Fuel Building Electric overhead Main hoist:
125 tons L202 Crane traveling bridge Auxiliary hoist: 25 tons 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 the hook center-line closer than 15 ft to the pool boundary, or by providing a suitable analysis demonstrating that for any failure mode, no heavy load can fall into the fucI-storage pool.
PGE Response:
The spent fuel pool (SFP) bridge crane can be excluded from the above category because this crane does not handle heavy loads.
The maximum load this crane will handle is a spent fuel assembly and its handling tool.
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 (ie, complete compliance with NUREG-0612, Section 5.1.6, or partial compliance supplemented by suitable alternative or additional designed features). For each crane so evaluated, provide the load handling system (ie, crane load combina-tion) information specified in Attachment 1.
PCE Response:
The Fuel Building crane has not been evaluated under the criteria of NUREG-0612, Section 5.1.6.
4.
For cranes identified in 2.2-1, above, not categorized according to 2.2-3, demonstrate that 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 spent fuel area and your determination of compliance.
This response should include the following information for each crane:
Which alternatives (eg, 2, 3, or 4) from those identified in a.
NUREG-0612, Section 5.1.2, have been selected?
Trojan Nucisar Plent Attachm:nt 1 Docket 50-344 Paga 2 of 3 License NPF-1 i
PGE Response:
Alternative 2 is selected.
b.
If Alternative 2 or 3 is selected, discuss the crane motion limitation imposed by electrical interlocks or mechanical stops and indicate the circumstances, if any, under which these protective devices may be bypassed or removed. Discuss any ad+1nistrative procedures invoked to ensure proper authorization of bypass or removal, and provide any related or proposed Technical Specification (operational and surveillance) provided to ensure the operability of such electrical interlocks or mechanical stops.
PGE Response:
The Fuel Building crane has mechanical stops and electrical interlocks which prevent movement of the hook centerline closer than 6 ft to the SFP. These protective devices may only be bypassed or removed while following an approved procedure, and then only with the Shif t Supervisor's approval. With the exception of the new fuel cask, movement of the hook centerline closer than 15 ft to the SFP is prevented by Plant procedures which require the handling of heavy loads over the safe load path on the 93 ft elevation of the Fuel Building, shown in Figure 1.
It is not considered credible that a procedural violation, a crane failure, and a load drop and roll of at least 6 ft or failure of the mechanical stops and electrical interlocks could occur simultaneously, resulting in a load drop into the SFP (this does not include a spent fuel cask). The new fuel cask is moved within 8 ft of the SFP in a horizontal configuration in order for the new fuel to be unloaded by the spent fuel pool bridge crane.
In this configuration, the risk of the cask falling into the SFP is considered extremely low.
When lifting a spent fuel shipping cask into the cask loading pit, the hook centerline will come within approximately 8.5 ft of the SFP. To preclude rolling if dropped, the spent fuel cask handling procedures require that the cask not be carried at a height higher than necessary and in to case greater than 6 in.
above the operating floor.
Since it is not known which cask will be used to ship spent fuel from the Trojan Nuclear Plant, it is not possible to verify that the criteria of Section 5.1 of NUREG-0612 are met for the spent fuel cask.
Before use of a spent fuel shipping cask, such verification would be performed under 10 CFR 50.59.
The requirement for this review prior to handling of a spent fuel cask will be incorporated in the appropriate Plant procedures.
Trojan Operating License, Appendix A, Technical Specifica-tion 3.9.7, requires that loads not be handled over the SFP which, if dropped, would have impact energies capable of pro-ducir; unacceptable damage to storage racks containing spent fuel. This requirement precludes heavy loads from being
Trojan Nuclocr Plant Attachasnt 1 Dockst 50-344 Page 3 of 3 License NPF carried over the SFP.
To prevent a potential drop of the Fuel Building crane main or auxiliary lower load blocks into the i
SFP, a design change will be made to install redundant upper travel limit switches on the main and auxiliary hoists to suffi-ciently reduce the risk of a failure from a two-blocking event.
As a result of these procedural and Technical Specification requirements, the inherent safety isatures of the Fuel Building crane and the expanded crane operator training and inspection and testing requirem.ents previously reported, it has been deter-mined that load handling operations in the vicinity of the SFP are in compliance with NUREG-0612.
Items 4.c, 4.d, and 4.e of Section 2.2 are not applicable to the PGE response.
s a
t SAB/4b1m3.lB4 l
Troj:n Nuclorr Plcnt Attcchm:nt 2 Dockst 50-344 Pign 1 of 3 License NPF-1 PGE'S RESPONSE TO INFORMATION REQUESTED BY THE NRC STAFF IN SECTION 2.3 0F ENCLOSURE 3 TO NRC LETTER OF DECEMBER 22, 1980 1.
Identify by name, type, capacity, and equipment designator, any crane physically capable (le, taking no credit for any inter-locks or operating procedures) of carrying heavy loads over the reactor vessel.
PCE Response:
Equipment Name Type Capacity Designator Containment Polar bridge Main hoist:
165 tons L201 polar crane Auxiliary hoist: 25 tons 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 f rom the movemeFt 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 land in or on the reactor vessel.
PGE Response:
The reactor cavity manipulator crane can be excluded from the above category because this crane does not handle heavy loads. The maximum load this crane will handle is a fuel assembly containing a rod cluster control element.
3.
Identify any cranes listed in 2.3-1, 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 (ie, 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 (ie, crane-load-combination) informa tion specified in Attachment 1.
PGE Response:
The Containment polar crane has not been evaluated under the criteria of NUREG-0612, Section 5.1.6.
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 of this request. With respect to Criteria I through III, provide a discussion of your evaluation of crane operation in the Containment and your determina-tion of compliance. This response should' include the following information for each crane:
1 Trojan Nucisar Plant Attschment 2 I
Docket 50-344 Pz6s 2 of 3 License NPF-1 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 authoriza-tion of such action. Discuss any related or proposed Technical Specification concerning the bypassing of such interlocks.
b.
Where reliance is placed on other, site-specific considerations (eg, refueling sequencing), provide present or proposed Technical Specifications and discuss administrative or physical controls provided to ensure the continued validity of such considerations.
c.
Analyses performed to demonstrate compliance with Criteria I through III should conform with the guidelines of NUREG-0612, Appendix A.
Justify any exceptions taken to these guidelines and provide the specific information requested in Attachment 2, 3, or 4, as appropriate for each analysis performed.
PCE Response:
In addition to establishing safe load-handling areas, training and qualifying crane operators and implementing increased inspection and maintenance requirements, load drop analyses have been performed to ensure safe load-handling operations in the vicinity of the reactor vessel. To sufficiently reduce the risk associated with the Containment polar crane main or auxiliary hoist lower load blocks, redundant upper travel limit switches will be installed to reduce the probability of a two-blocking event.
The heavy loads handled in the vicinity of the reactor vessel for which load drop analyses have been performed are the reactor vessel head, the core upper internals, and the missile shields.
The drop heights and locations used in the analyses are based on present operating practices for handling these loads. The general considerations of NUREG-0612, Appendix A were used in performing these analyses.
While removing the reactor vessel head, the potential exists to drop the head back onto the vessel. This drop could occur through air or a combination of air and water and could occur from various heights. In order to evaluate the separate cases, an analysis was performed using conventional energy balance methods similar to those described in Westinghouse WCAP-9198, Reactor Vessel Head Drop Analyses. These plant-specific analyses determined the consequences to the reactor vessel and its supports, and verified the conclusions reached in the Westinghouse report that "there will be no consequential damage to the structural integrity of the vessel nozzles and core-cooling capability, and the integrity of the fuel cladding will be maintained."
A drop of the missile shields from their maximum lift height was analyzed to determine the ef fects on the reactor vessel and head.
~
Trojen Nuclocr Plant Attschment 2 i
Dockat 50-344 Pzga 3 of 3 License NPF-1 Assuming a worst-cace orientation rf a missile shield on impact with the head, the analysis showed that penetration of the pres-sure bounda y would not occur. The consequences.to the reactor vessel and its supports would be less severe than discussed in the head drop analysis above, since the impact energies of the missile shields dropped from their maximum height are less than the energy of the worst-case reactor vessel head drop.
Analysis of a potential drop of the reactor vessel core upper internals into the reactor vessel was performed to determine if unacceptable damage would occur to fuel in the core or to the reactor vessel.
For a drop height of approximately 11-1/2 feet (when the internals lower plate clears the guide rods by about 6 in.), it was conservatively assumed the upper internals dropped through water to its original position in the vessel and the resulting impact energy was absorbed by the core barrel and the reactor vessel. The analysis demonstrated that sufficient strain energy absorption capacity exists in structural elements at contact interfaces such that the fuel will not be impacted and the reactor vessel will not be unacceptably damaged.
It is concluded that load handling operations in the vicinity of the reactor vessel are in compliance with NUREG-0612.
SAB/4blm3.lB8
Troj:n Nuc12cr Plant Attcchment 3 Docket 50-344 P:gn 1 of 5 License NPF-1 PGE'S RESPONSE TO INFORMATION REQUESTED BY THE NRC STAFF IN SECTION 2.4 0F ENCLOSURE 3 TO NRC LETTER OF DECEMBER 22, 1980 l.
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 extremely small for all loads to be carried and the basis for this evaluation (ie, complete compliance with NUREG-0612, Section 5.1.6, or partial compliance supplemented by suitable alternative or addi-tional design features). For each crane so evaluated, provide the lo9d-handling system (ie, crane-load-combination) information speci-fled in Attachment 1.
PCE Response:
The cranes identified in Section 2.1-1 have not been evaluated under the criteria of NUREG-0612, Section 5.1.6.
2.
For any cranes identified in 2.1-1 not designated as single-failure-proof in 2.4-1, a comprehensive hazard evaluation should be provided which includes the following information:
a.
The presentation in a matrix format of all heavy loads and poten-tial impact areas where damage might occur to safety-related equipment. Heavy loads identification should include designation and weight 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 impact area can be located on the plant general arrangement drawings.
PGE Response:
The requested information is provided in Tables 1 through 4.
b.
For each interaction f dentified, indicate which of the load and 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.
PGE Response:
No interactions identified in Tables 1 through 4 were eliminated from consideration because of separation and redundancy of safety-related equipment, mechanical stops, electrical interlocks or other, site specific considerations.
For interactions not eliminated by the analysis of 2.4-2-b, above, c.
identify any handling system for specific loads which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely small and the basis for this evaluation (ie, 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,
Troj n Nuclssr Plant Atttchm2nt 3 Dockat 50-344 Paga 2 of 5 License NPF-1 provide the load-handling system (ie, crane load combination) information specified in Attachment 1.
PGE Response:
No interactions idantified in Tables 1 through 4 are eliminated from consideration because the likelihood of a load drop is extremely small as defined by NUREG-0612 criteria. The handling syster design features have not been evaluated under the criteria of NUREG-0612, Section 5.1.6.
d.
For interactions not eliminated in 2.4-2-b or 2.4-2 c, above, demonstrate using appropriate analysis that damage would not pre-clude operation of sufficient equipment to allow the system to perform its safety function following a load drop (NUREG-0612, Section 5.1, criterion lV). For each analysis so conducted, the following information should be provided:
- 1) An indication of whether or not, for the specific load being investigated, the overhead 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 guide-lines of NUREG-0612, Appendix A.
- 3) The information requested in Attachment 4.
PCE Response:
The interaction identified in Table 1 for the Auxiliary Building Equipment Removal Hoist will be eliminated f rom further considera-tion by a design change either to move the HVAC duct which protrudes into the hoistway or t9 shield the safety-related cable trays from damage as a result of the HVAC duct failure. The train "B" essential cable trays are not otherwise susceptible to a potential load drop in this hoistway.
The interactions identified in Tables 2, 3 and 4 are eliminated from further consideration by structural analyses which demon-strate the acceptability of a load drop.
In each location, load drops were analyzed based on present operating conditions and the most limiting load handled over that location. Where the conse-quences of a load drop were shown to be unacceptable, mitigating actions or changes to operating procedures will be implemented which will reduce the severity of the drop to an acceptable level. The general considerations of NUREG-0612, Appendix A, were used in performing these analyses.
For the interactions identified for the Turbine Building crane in Table 2, a load drop analysis was performed for the generator rotor which is the most limiting load handled over the potential
Tt;jsn Nuclear Plant Atttchmant 3 Docket 50-344 Pagn 3 of 5 License NPF-1 target area.
It was considered that the "B" emergency diesel generator on the 45-f t elevation must be protected frc= this Joad drop. The load drop is assumed to occur from a height of 3 ft immediately af ter the 210-ton rotor is withdrawn from the generator.
The rotor would first impact the Turbine Building operating floor, a 12-in. concrete slab supported by steel framing. Beneath the operating floor is the roof of the "A" ESF switchgear room, a 6-in. concrete slab supported by steel framing. The "B" emergency diesel generator is also protected by the east end of the floor of the "A" ESF switchgear room.
The analysis showed that the structural response was acceptable for a drop of 6 in. or less.
If the rotor is being transported across the operating floor at a height of 6 in., some restric-tions on the load path are also required when above the ~B" Diesel Generator Room.
As a result, when the rotor is being withdrawn from the generator, actions will be taken to reduce the effective drop height to 6 in. or less. Mitigating actions and/or changas to operating procedures are being considered to prevent the 3-ft height free drop of the rotor. In addition, if the rotor is to be transported above the "B" diesel generator room, administrative action will be taken to establish the safe load path. All other heavy loads handled by the Turbine Building crane may be lifted safely over the floor area under consideration at a height of 6 in. or less.
Load drop analyses were also performed for the Fuel Building crane interactions identified in Table 3.
With the exception of the spent fuel shipping cask, which is not addressed in this response, the reactor coolant pump motor is the most limiting load that could be handled over the potential target areas in the Fuel Building. The conditions / assumptions used for these analyses are representative of the actual conditions for a movement of the reactor coolant pump motor from the equipment hatch service rail through the Fuel Building along the designated safe load path.
The pump motor was analyzed for a drop of 6 ft in the vicinity of the equipment hatch service rail onto the operating floor of the Fuel Building which is a 30-in. reinforced concrete slab supported by steel framing. For certain potential drop locations near the end of the service rail, the analysis has shown that structural response due to a 6-ft drop is acceptable. A 6-in. drop has been shown to be acceptable along the entire length of the safe load path except that the pump motor should not be handled over the sections of the operating floor which consist of 8-in. con-crete slabs. The result of these analyses will be incorporated in operating procedures for RCP motor removal when these proce-dures are written. For this lift, the safe load path will be marked such that the 8-in. slabs will be avoided.
An analysis to determine the consequences of a potential drop of the reactor coolant pump motor on the 93-f t elevation inside
Troj:n Nuclocr P]cnt Attichm2nt 3 Dockat 50-344 Page 4 of 5 License NPF-1 Containment was also performed. This interaction is listed in Table 4 with the potential targets being the train "A" and "B" residual heat removal (RHR) suction and discharge piping near Elevation 60 ft.
The results of this analysis demonstrates that the consequences of the load drop are acceptable and that no mitigating actions or procedural changes need to be implecented.
Analyses were also conducted for the potential drop of the reactor vessel head onto the 93-ft elevation concrete slab and onto the reactor vessel head storage stand. The potencial targets for this drop are the train "A" and "B" RHR suction and discharge piping near Elevation 60 ft.
A drop of the reactor vessel head onto the concrete slab is acceptable from a height of 2.5 ft or less. This height limitation will be implemented in the appcopriate Plant procedures. The analysis of the head drop onto the storage stand is incomplete. However, as a result of the analytical work done thus far, design changes have been identified which, if implemented, would obviate the need for further analysis.
If the results of the analysis and associated handling procedure changes do not demonstrate the acceptability of a potential drop of the reactor vessel head onto the storage stand, a design change option will be implemented.
A drop of a reactor coolant pump motor onto a reactor coolant loop during Mode 5 with a resultant double-ended break was also evaluated. It was determined that the consequences of this accident were acceptable and were less severe than the case analyzed in the Final Safety Analysis Report.
The Technical Specifications require suf ficient equipment to be operable in this mode to adequately provide continued decay heat removal.
The seismic design for the Fuel Building, Containment polar and Turbine Building cranes are described in the crane design speci-fications. The specifications for the Fuel Building crane and the Containment polar crane include requirements that these cranes be designed to resist the most severe of the following combinations:
- 1) Crane gravity loads plus simultaneously applied horizontal and vertical OBE inertia loads, with stresses limited to 125% of normal allowable working stresses, or
- 2) Crane gravity loads plus simultaneously applied horizontal and vertical SSE inertia loads, with stresses limited to 150% of normal allowable working stresses (not to exceed 0.9 F in tension or flexure, or 0.5 F in shear).
y y
Horizontal inertia loads were based on the weight of the crane excluding the lifted load, and vertical inertia loads were based on the weight of the crane including the lifted load.
The Turbine Building crane specification included requirements that the crane be designed to resist seismic loads as defined by the Uniform Building Code for seismic zone 2.
Trojan Nucisar Plant Attachssnt 3
^ Docket 50-344 Pags 5 of 5 License NPF-1 As a result of these analyses and the actions taken to modify the load handling practices where necessary, it is concluded that load handling operations in the vicinity of safe shutdown or decay heat removal equipment are in compliance with NUREG-0612.
s l-l i
e i
d SAB/4bim3.1B24-L
Trojan Nucic:r Plcat Attrchm:nt 4 Dockat 50-344 P gs 1 of 1 License NPF-1 PGE'S UPDATE OF THE INFORMATION PROVIDED IN OUR LETTER OF SEPTEMBER 22, 1981 IN RESPONSE TO SECTION 2.1.3-d of ENCLOSURE 3 TO NRC LETTER OF DECEMBER 22, 1980 3.d.
Verification that lif ting devices identified in 2.1.3-c comply with the requirements of ANSI N14.6-1978, or ANSI E30.9-1971, as appropri-ate.
For lif ting devices where these standards, as supplemented by NUREG-0612, Section 5.1.1 (4) or Section 5.1.1 (5), are not met, describe any proposed alternatives and demonstrate their equivalency in terms of load-handling reliability.
PCE Response:
The PCE response of September 22, 1981 indicated that a program would be developed to meet the continued testing, maintenance and repair, and nondestructive testing requirements of Sections 5.3, 5.4 and 5.5 of ANSI N14.6-1978.
Further consideration of the applicability of the testing requirements of Section 5.3, the ALARA concerns of performing these tests and the results of load drop analyses has led to the determination that annual testing is not justified.
The annual testing requirements for special lifting devices imposed by ANSI N14.6-1978 are not justified since the frequency of use, and therefore the wear and tear, of the special lif ting devices for the reactor vessel head and core internals is much lower than for shipping containers for which the standard was written. In addition, the special lifting device for the core internals is painted, and highly contaminated from being submerged in the refuelirg cavity. Therefore, the testing and inspection in accordance with ANSI N14.6-1978, Section 5.3.3.2, is not justified as it could result in an exposure of 5-10 man-rem. Finally, since the consequences of a load drop of the reactor vessel head or core upper internals has been demonstrated to be acceptable, implementation of the continued testing requirements of ANSI N14.6-1978 is not considered necessary. The present practice of conducting a visual inspection of the lifting devices prior to each lift will be continued.
The PGE response of September 22, 1981 also deferred addressing the applicability of Section 6 of ANSI N14.6-1978 to the reactor vessel 1
head and core opper internals. Thic deferral was based on obtaining the results of load drop analyses to determine if either the reactor vessel head and/or core upper internals should be considered a cri-tical load. A critical load, as defined in ANSI N14.6-1978, is "any load whose uncontrolled movement or release could adversely affect any safety-related system when such system is required for unit safety or could result in potential offsite exposures comparable to the guideline exposures outlined in Code of Federal Regulations, Title 10, Part 100".
As a result of load drop analyses conducted for the reactor vessel head and core upper internals (included in this response), neither load fulfills the definition of a critical load, and therefore the requirements of Section 6 of ANSI N14.6-1978 are not applicable.
SAB/4blm3.1B12
i> :'.'
[
i i4 f
If iIlti <i 1
n o
i t y ar d no rig ame zit al a HEC 4
tn
- d e yem t t p eai fl u aeq SRE no i
t t,
ave lE no i
t y ar d no T
rig S
ame I
zit O
al a r
H HEC L
A V
t O
n M
- de E
yem R
t t p
. T fl u eai N
aeq E
SRE M
P 1
IU n
E Q
o L
E i
B t
A G
a T
N v
I e
D l
L E
I UB n
T o
R i
A t y I
ar 1
L dno I
rig K
9 ame U
4 zit e
A gF -
al a g
n 6
HEC n
i E4 a
d h
E l
c
~
N i s :
t "B
la n
A u as n
R B ee
- d e
" i g
C nn yem t
i yii t t p nnes s
rLL eai i el y e
a fl u asb a d
i nn aeq rsar l mm SRE Tect y
i uu b
xll u oo d
A CC n
e o
t i
t a
t f
n a
f
~
v 5
e i
e l
l E
e
~
e b
ae l
r l
i A
1 w
tc f,
d a
sn o2 r
p ai 2
a 7
m f r z
1 I
doer a
8 e
t e h
3 sb1 tb n
ui em s
i i
b s
orel e d
i cl t1 h
4 a
rsbE p8 T
/
o aeaC e9 B
L VdTPS1 1
A S
a; il
P TABLE 2 CRANE: TL1tBINE BUILDING c
Impact Area Turbine Building Column Lines:
S'-W' imeds Column Lines: 45-52 Safety-Hazard Safety-Hazard Safety-Hasard Related F.11mination Related Elimination Related Elimination Elevation Equipment Category Elevation Equipment Category Elevation Equipment Category Va rious - as 93 f t described in Table 2 of 63 f t "A"
ESF e.
PGE letter of switchgear September 22, room e.2
^
1981 45 f t "B" Train emergency diesel generatore e
1 The lopect area for the dropped loada would be on the 93-ft elevation operating floor. For damage to w cur to the safety-related equipment listed, the load would have to penetrate the concrete slab which is the operating floor and the concrete slab which is the roof of the "A" ESF switchgear room and impact the floor of the "A* ESF switchgear room with sufficient force to damage the "B" diesel generator.
2 Load drop ansylsis shows that the load will not penetrate the operating floor of the Turbine Building. In some cases, mitigating actions or procedural requirements are in place to limit the impact energy of the dropped load.
These Jitigating actiona and procedural requirements are discussaed in other sections of this report.
SAS/4b133818
s 4
. s n
o i
t y ar d no rig ame zit al a HEC tn de yem t t p y
eai r
fl u a
aeq d
g SRE noce no r
s, e
i Iu t
ta a
v e
e r
l c
E ro c
n sn o
b o i
ai t y l t ar sa d no v
rig tlee ame rit ee l a r
bEC cc nf vI o-c5 t
4 n
n
- de oe yem s
ih G
t t p t t N
eai a
I fl u
- vn D
aeq eo L
SRE l
I et U
n 3
B t e n
f m E
L o
- p L
E i
1i B
U t
6 u A
F a
q T
v d e e
n l
ad E
E e
N tt A
f a R
l C
n 7 e o
7 r -
i 4
,y y
t r
11g tt h
d oo f e t
rig
- - - e f
a ase 3 a p
zit 9 s 0
al a 6d HEC ee D - a hh 5 o1 t t C5l e
t t
d et er n
n n
t c g
- f u d e eg as aa ns sag yem nn r
rp 1
ieesi t t p oirse t m d nn F
eai pl e pl ei lii g fl u mot mo n
iLLnn aeq ooauo eo u
oi SRE ccwpc pt B nnl mman os l uu w
t n elld o n
o uoonh o
- n. i FCCas i
tttt t
t ffff
. a a
hv v
3715 e
e 9764 dl l
l e E
uoe wse a
dh e
at b
r o
A l mo t
d r c
ef a
p 9
p ps 1
a s e 8
l r
rl 3
os di n
t n s
i oo es b
s Mt hi 4
d T m 88 a
C4 1
A L
R S
O l'
I, 2
.l
s f
.Y4 f
I e
71 I
n o
i t y ar d no rig ame zit al a HEC tn de yem t t p eai t
fl u c
aeq a
SRE p
m i
e.
n o
o t
i t
s a
e v
l e
i l
s E
s i
m n
y o
r i
a
?
t y d
ar n
d no o
rig e
c ame e
zit s
al a HEC e
t s
n u
e a
m t
c E
ns n
N i u
- d e rt r
A ao yem ong o
R ti t t p t an C
nr eai cli b
oa fl u aop a
R CV aeq eoi l
A SRE R cp a
LO e
4 P
t n
e E
T o
r L
f i
t c
r B
E t
f n
A M
a o
T N
v 5
c I
e 4
A l
n T
E o
i N
t O
C a
n v
o e
i l
E t y y,
e.n N
ar I.
t o A
d no R
rig
- e
- e fi C
6 ame
- t 3 a 6
zit H -
al a 9 v 0
HEC e
E6 el h e m
m t
t ne ne t
ss n
it it ef t ee
- d e d as d as t -
nnn yem nry nry a0 eii t t p
- aTS
- aTS r6 t
mLL eal R
R ee n
fl u "A "B H "A "B H nh i nn aec amm SRE
" " R
" " R et p
t uu n
nll ooo oo t
CCC n
t o
i t t tt en t
ff ff ve am av 30 30 h p i
e 96 96 d u l
l q E
ue owd a
e e
e n
dt r
o aa ol A
s t
l e n
t l o 3
r c
et 4
d -
a s
ey p
s6 pt 0
m e4 pe 2
I V1 r
of 8
o ra 3
r-t d s m
1 o
o s
t d M
ee b
hh 4
d ca a
ae P
Tt
/
eH C
B o,
R R
1 A
~
S 1
J i
2 I
)'
e 9
9 9
o, 9
c_= _= ~__2 e, p.Ts~~ ~ [- d.
~*~
P+ r- -h b, o.
5
@"~
d,< m h '..
,7 f~%' ii,[
W
- v -
~
Ip g*
W. ',.r f Id g 7. q<,y $,yc_-m-- T..,, *,*Jy"'<Dq
_e,7.. l.
3'3 J
1@
.t.
i
! t l
u
.. -. [:J_, rtis8?s ar'
- y ";_
l%
j.
~
.4-
- N-y Q,,
1%
@.-A LC
. ? "- 'T I
,.re.
g#_ -
I "j, 7b e s;",;
F
/
./ U., '
I
"*:'}.a, l
y, L
s
~~
~
g
? h i
t
'Y l
erv.a2a. s p
--A
.I r
6..
---.J V
y' -
"t. '
.- {i
/\\
(1 l;
[
I I
_ -,-4 an.; s._...,,
JU.
_. 4}q
[
. f -e n ===.*,
g sw r Y y
' *-g,- i 4
_,.L' r
u g
. 4t's='= *. 24_..
j
+-- _
k r
L 7d
-- r t'"d *-
1
-g l
PLAN t.T Et T rf I
l
- -'~5
[
p l (',,ajf~
w._] u..,.
ps t
u.
,,a,
.=.
m., - w.r A.. b t r ;*
-y y -,. +
2,
~.
1 ix.
,u-..
- I
,.j,
g I
I
- -** was a
(
,.3 t,_a ~--
m=---
- r
-v.-
. y-n._%,,,,
-H<--------
~~
-- p
__ht
'i
~
i.. '.. r '
.I 1,
t*
m.~.
p,,}*:
e
- e,..t e'em Ld h
- I
.L, -- --
. _. 2__..-
- ._. f_se
- h..
b
, h,
m...
,,,,,%,,g o.,.
r a.
i L.....+-
w lt V M(" M_^' *.,'
l C
r
?n i
~
%..e s.
r
,}
,. I, u
'j l' "
j I'
I i
o v
i; pp d
a 7
i I'
L.
i.
I g-
^y
_Pz L
-L; j
g WJ
__ b gar-=- g K eQ74 L
.4 PL AN A' El 77D
^
i
~DfA~_ -.
- 4..m&
-.J 4.,
--.A..
k_.
.(
@-_.,/....,--g..
y g.
s m
~ &-
+
.. ~ -
3
/:,. a.
3,. i,/
..%.'/?,....b ' ' **
.=r _.
.q
+
A,,..b.,' "* v..,
(*
C,;.r ***
m
\\ ~
v_ ;-. vi._._ _
Z:
i 4 ~..._...
1,
._ h.f a a e:".-
e y.
3.__
l r l.,...
d
- _, ;
- D m
Q
.,g
/'"-- ~s
."m.-.'.,
,/.
4 b
_ i H.
/ 'A*-,
!_ L,'";'".l.";
._L_
v w,,
>+. i.~.~
E i
p._....>2
. -.1 1
_._-+
l.
p i-,a 6
V -
f,y;,..xq,.m.
- ,, p
+
Q ;
.w
)
fi,.,
. e m.a r.
l t......,$,~ m p q lf _.1 g
r :w e x*
.m s>%
a t.
.n...
i. - - - - -. -
l a,
T b ~
.J. I.
t_..
m Q.-
=m r.-
-. y
..m
~
, y PL AN A' Et Oi -C' A
P1
,_ =
e.g--
f me-w-7,,
L Qw,. - __ - _ - -
- ~ -.ll f.
l
_4. h ' l* -y-y:-
s d---
- k. e.,.
' ~ "__ % < M, F " l ii
.i n
m e
k J l j [ ;(w
-4
~
i.=., kJ l
a' 1.i [..
Vf
, ~,..,
,+
T,
,,,c_,.
,;.,m i
v n._x._,
- v... m- --
pwm u.i.
j1 c
g 3.
g<
_-t
.. is.%. ~
- E l
g y-r.
a..~
- u. j.
,1 ;l l
t
~ tr - ~ ~. e *=e. -- - - O ;,,_. se... l 4
11 g_
g._ j
-.i--
j
.i o-we
_o __ _ __
l
+
l
......{.[%- G ~
8 5.=*.. _.
__n.
s
..e, Q ;, W.. ',.
l "ll_
l l
i 1
.l :,,
}
.i g
- .e.
_ J" 2,g,'*d,*
f.
f,,,__.
~.
. {..
iL__
t; h
.h
..t 2
4
__k '"-
6)
J 4
At AN At Et4Q O
e 6
u,.u.- - ~ w. m,u m m FIGUpt l PLANT ArmeggutNT DIAQaw or ars :r wom e :umpt' 3. uon umimis M. W -
FUf L CASK wovtWENT E*ntLCPE Amendment 5 (October 1973)
.