W3P81-2158, Forwards Part II of Rept Re Overhead Handling Sys Operating in Spent Fuel Storage Pool,Containment & Inplant Areas Containing Reactor Shutdown Equipment.Load Drop Analyses Results Will Be Available by June 1982
| ML20031A369 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 09/21/1981 |
| From: | Maurin L LOUISIANA POWER & LIGHT CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0612, RTR-NUREG-612 3-A1.01.04, 3-A34.14, W3P81-2158, NUDOCS 8109230408 | |
| Download: ML20031A369 (15) | |
Text
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LOUISIANA f 2 e ouinoNot sinur POWE R & LIGH T/ p o Box 6008
- NEW OnLEANS. LOU 19Akh 70174 * (504) 3642345
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L V. MAURIN Asst. Vice President Nuclear Operations September 21, 1981 W3P81-2158 3-A1.01.04 i/d'r 3-A34.T40 s
.,G' g
'ek Mr. Darrel G. Eisenhut S-h(. b Directur. Division of Licensing 2
N y g' U. S. Nuclear Regulatory Commission g
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Washington, D. C.
20555
SUBJECT:
Waterford 3 SES
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Control of Heavy Loads NUREG-0612
REFERENCES:
(1) Letter from D. G. Eisenhut dated December 22, 1980 (2) Letter W3P81-1513 dated June 19, 1981
Dear Mr. Eisenhut:
Reference (1) transmitted a copy of NUREG-0612 and requested that we evaluate the control of heavy loads utilizing overherd cranes. Reference (2) transmitted Part I of a report which covers the get eral requirements for overhead handling systems at Waterford 3.
Enclosed with this letter is a copy of Part II of the report which covers specific requirements for overhead handling systems operating in the vicinity of the spent fuel storage pool, in the containment, and in plant areas containing equipment required for reactor shutdown, core decay heat removal, or spent fuel cooling. We have committed to performing several load drop analyses in the report, and the results of these analyses will be available by the end of June, 1982.
Yours very truly, 77Netu,;
L. V. Mauria LVM/RMW/pjl 3o30 cc:
E. Blake, S. Black, W. M. Stevenson Ifl 6109230408 810921 PDR ADOCK 05000382
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INTRODUCTION 4
' The purpose of this report is to review and evaluate the design and operation of overhead handling systems in accordance with the guidelir.es of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants", and.NRC's " Request for Additional Information on-Control of Heavy Loads".
1 l
This. report is submitted in two parts.
Part I covers general requirements for overhead handling systems.
Part II covers specific requirements for overhead handling systems operating in the vicinity of the spent fuel storage pool, in the contain-ment and in plant-areas containing equipment required for reactor j
shutdown, core decay heat removal, or spent-fuel pool cooling.
The objectives for the control of heavy loads, the defense-in-i depth approach for controlling the handling of heavy loads, the extent of the overhead handling system under review, the identi-fication of the heavy loads as well as the general outlines of the evaluation have been presented in Part I of the report.
Part j
II is to further expand the evaluation and provide more specific information as identified in Sections 2.2, 2 3 and 2.4 of NRC's i~
" Request for Additional Information on Control of Heavy Loads".
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a 1
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1.
SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN THE VICINITY OF FUEL STORAGE POOLS The information provided below is generally in accordance with the guidelines concerning the design and operation of load-handling systems in the vicinity of stored spent fuel as given in NUREG-0612, Section 5 1.2.
1.1 CRANES CAPABLE OF TRAVELING OVER THE SPENT FUEL STORAGE POOL The following cranes are capable of traveling over the spent fuel storage pool:
a)
Fuel Handling Building Bridge Crane - this crane consists of one main heist of 125 ton capacity and two auxiliary hoists of 15 ton capacity each.
This crane is used to handle the spent fuel shipping cask, pool gates, hatch covers, new fuel containers, plant equipment and all necessary miscellaneous operations.
b)
Spent Fuel Handling Machine - this machine is also a bridge and trolley type handling equipment equipped with electrical interlot':s for both hoist and trans-lation movements.
This machine is used to h ndle the fuel assembly exclusively.
1.2 EXCLUSION OF SPENT FUEL HANDLING MACHINE The Spent Fuel Hand 71ng Machine is used for moving fuel assemblies and is generally not used for handling of any heavy loads.
Therefore, it is excluded from the evalua-tion in this report.
1.3 SINGLE - FAILURE - PROOF The Fuel Handling Building Bridge Crane is not considered as single-failure-proof, as it does not comply completely with NUREG-0554, " Single-Failure-Proof Cranes for Nuclear Power Plants" as called for in NUREG-0612, Section 5.1.6.
To upgrade the existing crane to improve the reliability of the handling system is also not contemplated.
1.4 EVALUATION OF CRANE DESIGN AND OPERATION IN THE SPENT FUEL ARES a)
To previde assurance that the evaluation criteria of NUREG-0612, Section 5 1 are met in addition to satis-fying the general guidelines of NUREG-0612, Section 5.1.1; the alternative (2) as identified in NUREG-0612, Section 5 1.2 has been selected. _ _ _ _ _ _ _ _ _ - _ - _ _ - _ _ _ __ __ _
b)
The Fuel Handling Buidling Bridge trane is provided with an electrical interlock system.
The restricted zone limit switch devices and logic will preclude crane entrance to the spent fuel area pending loca-tion of the trolley, so that the main hook cannot carry a spent fuel cask or any other heavy loads over the spent fuel storage pool or the pool walls.
The electrical interlock will be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereafter during crane opera-tion (see FSAR Chapter 16, S':; tion 4.9.7 for surveillance requirements).
There is.no anticipated maintenance in the spent fuel pool area requiring the use of the main hook; however, if needed, the electrical interlock can be by-passed or removed, but this will be controlled by an administrative procedure to ensure adequate protection.
The auxiliary hooks are permitted to travel over the spent fuel storage pool when servicing the poison inserts (service weight is approximately 200 pounds including the tools) during surveillance teet.
This crane prohibits loads in excess of 2,000 pounds from travel over fuel assemblies in the pool.
This will be included in an administrative procedure.
In addition, an administrative control will also be exe: -
cisea during crane operation (see FSAR Chapter 16, Sections 3 9 7 and 4.9.7 for limiting condition for operation and surveillance requirements respectively).
NUREG-0612, Section 5.1.2 (2) (a) calls for mechanical stops or electrica_ interlocks to be provided that prevent movement of the overhead crane load block over or within 15 feet horizontal of the spent fuel pool.
According to the plant arrangement, the curren' main hook limits allow the load block to travel closer to the pool - 7 feet 6 inches from north side of the pool and 7 feet 61 inches from west side of the pool, or 9 feet 6 inches from northernmost fuel assemblies and 10 feet 7 inches from westernmost fuel assemblies.
The plant arrangement also indicates that the spent fuel shipping cask is exclusively placed in its own storage and decontamination pools which are indepen-dently supported on a huge mass concrete slab and are also completely separated from the spent fuel storage p001 by a three foot thick concrete wall extending to the operationg floor (elevation +46 feet).
In addition, the Fu21 Handling Bu* Ading Bridge Crane is also provided with a relundant geared limit switch arrangement with a key lock bypati on the main push- _ _
button penuant limiting the high point of the main hook to elevatation +58.5 feet.
This further pre-cludes any possibility of lifting the cask so high that it might drop and roll over the partition wall and fall into the spent fuel storage pool.
Figure 1 demonstrates the cask handling conditions.
Under no circumstances can the e.. tire cask be carried at a height higher than the operating floor level.
c)
No reliance is placed on crane operational limita-tions with respect to the time of the storage of certain quantities of spent fuel at specific post-irradiation decay times.
d)
No reliance is placed on the physical location of specific fuel modules at certain post-irradiation decay times.
e)
Based upon the crane featrues, load and lifting limitations as well as plant arrangements as dis-cussed in this part of the report, Section 1.4(b);
it is concluded that the likelihood of a load drop, which nay damage spent fuel, spent fuel stcrage racks and spent fuel storage pool, is extremely small.
This, in turn, assures that it will not result in any radioactive material release or unacceptable fuel configuration or pool water leakage.
Therefore, further analyses to demonstrate compliance with the evaluation Criteria I through III are not required.
To determine any consequen; effects on the strut bure following accidental dropping of a postulated heavy load, the following load drop analyses will be per-formed as indicated in Part I of the report:
1)
Gates #3A and #3B to be dropped on their storage area bottom slab 11)
Crane main hook load block to be dropped on the operating floor The above analyses are scheduled to be completed by the end of June, 1982.
I 2.
SPECIFIC REQUIREMENTS OF OVERHEAD HANDLING SYSTEMS OPERATING IN THE CONTAINMENT.
'Information provided'below are generally in accordance with' thedguidelines concering the design and operation of load-handling systems in the vicinity of the reactor core as given in NUREG-0612, Section 5 1 3 2.1.
CRANES CAPABLE OF TRAVELING OVER THE REACTOR VESSEL The following cranes are capable of traveling over the.
reactor vetael:
a)--
Polar Crane - This crane consists of one main hoist of 200 ton capacity and one auxiliary.
hoist of 30 ton capacity.. This crane is used to handle the reactor vessel head, upper. vessel internals, reactor coolant pump motors, plant equipment and all necessary miscellaneous operations.
i b)
Refueling Machine - This machine is also a bridge
~
and trolley type handling eugipmect equipped with both mechanical stops and electrical interlocks for hoist and translation movements.
This machine is used to handle the fuel assembly and control element assembly exclusively.
2.2 EXCLUSION OF REFUELING MACHINE
.The refueling machine is used for moving fuel assemblies 4
and control element assembliec, and is generally ~not used for handling of any heavy loads.
Therefore, it is ex-cluded from the evaluation in this report.
23 SINGLE - FAILURE - PROOF
.The polar crane is not considered as single-failure-proof, it does not comply completely with NUREG-0554, " Single-Failure-Proof Cranes for Nuclear-Power Plants" as called for in NUREG-0612, Section 5 1.6.
To upgrade the existing crane to improve the reliability.
of theLhandling~ system is also not contemplated.
2.4 EVALUATION OF CRANE OPERATION IN THE RELCTOR VESSEL AREA To provide assurance that the evaluation criteria of NUREG-0612,;Section' 5.1 are met in addition to satisfying the
~
general guidelines of NUREG-0612, Section 5.1.1; alternative (2) as identified;in-NUREG-0612, dection 5 1.3 has been selected. i i.
a
Tc satisfy the evaluation criterion I, a rapid containment isolation system has been provided with prompt automatic actuation on high radiation.
FSAR Sections 6.2.4, 7.6.1.5 and 9.4.5 3 describe that the area radiation monitors located inside the containment will generate a. Containment Purge Isolation Signal (CPIS) upon detection of radioactivity above a set point.
The CPIS closes the containment pu.ge isolation valves regardless of plant operating mode.
This action will prevent release of containment air which-contains an unacceptable I
level of. radioactivity.
This system is composed of two redundant safety channels, they are physically and electri-cally separated.
Testing will also be performed periodically as outlined in FSAR Chapter 14.
The above assures that postulated releases are within limits of evaluation criterion I.
To satisfy evaluation criteria II and III, a po3tulated reactor vessel head drop analysis will be performed.
This analysis is scheduled to be completed by the end of June, 1982.
a)
As described in Part I of this report, this polar crane is provided with mechanical stops to restrict the crane movement within the hook limits, and electrical interlocks to restrict the hoist motion to its upper and'1ower limits.
No other mechanical stops or electrical interlocks are provided to eli-mina'e the heavy load handling operation over the c
reactor vessel and irradiated fuel assemblies.
b)
No reliance is placed on other site-specific consi-derations.
c)
Based upon the evaluation of crane design and i
operation described above, it is. concluded that a rapid containment isolation system has been pro-vided, and no further analysis is required to demonstrate compliance with' Criterion I.
- However, a postulated reactor vessel head drop analysis will l
be required to demonstrate compliance with Criteria II and III.
To determine any consequent effects on the structure following accidental dropping of a postualted heavy load, the following load drop analyses will also be performed as indicated in Part I of the report:
1)
Reactor Core barrel to.be dropped on the reactor canal bottom-f 11)
Reactor. coolant pump motor to be dropped on the' operating floor or floors below 6-m.
- m. m
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iii)
Plant equipment such as containment fan cooling element to be dropped on the operating floor All postulated heavy load drop analys'., are scheduled to be completed by the end of June, 1982.
P 3
SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN PLANT AREAS CONTAINING EQUIPMENT REQUIRED FOR REACTOR SHUTDOWN, CORE DECAY HEAT REMOVAL, OR SPENT FUEL POOL COOLING The information provided below is generally in accordance with the guidelines concerning the design and operation of 1 cad-handling systems in the vicinity of equipment or components required for safe reactor shutdown and decay heat removal as given in NUREG-0612, Section 5 1 5 The overhead cranes and their heavy loads which are evaluated are identified in Part I of the report, Sections 3 and 4, respectively.
The safe load travel paths are defined in Part I of the report, Section 5 3 To provide assurance that the evaluation criterion IV of NUREG-0612, Section 5.1 is met in addition to satisfying the general guidelines of NUREG-0612, Section 5 1.1; alternative (1) (c) as identified in NUREG-0612, Section 5.1 5 has been selected.
Where the safe shutdown equipment has a ceiling separating it from an overhead crane, alternative (2) as identified in NUREG-0612, Section 5.1 5 will be used-to show by load drop analysis that the postulated heavy load would not penetrate the ceiling or cause spalling that could cause l
i failure of the~ safe shutdown equipment.
The required load drop analyses to be performed have_been identified in Sections 1.4(e) and 2.4(c) of this part of the report for Fuel Handling Building bridge crane and Reactor Containment Building polar crane, respectively.
For Reactor Auxiliary Building radwaste cask handling bridge crane, a load drop analysis for the crane load block will also be performed to determine any consequent effects on the floor below.
All load drop analyses are scheduled to be completed by the end of June, 1982.
3.1 SINGLE-FAILURE-PROOF The cranes listed in Part I of the report, Section 3 are i
not considered as single-failure-proof nor as in complete
[
compliance with NUREG-0554, " Single-Failure-Proof Cranes L
for Nuclear Power Plants" as called for in NUREG-0612,-
Section 5 1.6.
To upgrade the existing crancs to improve the reliability q
of the handling system is also not contemplated. k i
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32 COMPREHENSIVE HAZARD EVALUATION a)
The Reactor Containment Building (RCB) polar crane will be only operated during refueling or mainte-nance periods, not du"ing plant operation, nor after LOCA.
The Fuel Landling Building (FHB) rad-waste cask handling bridge crane can be operated all-times.
As identified in Part I of the report, there is no heavy load which, handled by either the FHB bridge crane or the RAB radwaste cask handling bridge crane,-might damage safety-related equipment-on the floor directly underneath any defined crane travel paths.
As for the RCB polar crane, there is only one heavy load No. 11 (200T main hook load block) which may travel near to two (2) vertical duct systems.
The latter systems are necessary to continually remove the residue heat inside the containment and main-tain the temperature even during refueling or maintenance.
The two duct systems are physically.
separated, so no load drop can damage both ducts at the same time.. Assuming one of the ducts.is damaged, the containment temperature will increase; however, this will by no means result in any radio-active material release,. damage to fuel, reactor-vessel or any other safe shutdown equipment.
It j
is therefore concluded that_this can be tolerated E
without violating the evaluation criteria of NUREG-0612.
There are also two (2) radiation monitors located on the operating floor inside the containment.
For evaluation, see Section 2.4 of this part of the report.
Table 1 presents a matrix format of the heavy load and impact area where damage might occur to safety-related equipment.
b)
In Table 1, the hazard elimination category'is also indicated for each interaction identified.
Any elimination is further supplemented by the following specific information:
(1)
For load / target. combinations eliminated because of separaticn or redundance off safety-related. equipment, discussiens have been given in Section 3 2(a) above.
l -.
(2)
No particular mechanical stops or electrical interlocks are provided to climinate any load / target combinations.
(3)
No other site-specific considerations are taken to eliminate any load / target combina-tions.
c)
All interar,tions are eliminated by a hazard evaluation as shown in Section 3.2(a) and 3 2(b) above.
d)
Same as (c) above.
SUMMARY
As a follow-up of Part I of the report, specific requirements for overhead handling ssstems operating in1the vicinity of spent fuel storage pool, in the containment and in plant areas containing equipment required for reactor shutdown, core decay heat removal or spent fuel pool cooling, are given in this Part II of the report in accordance with the guidelines of NUREG-0612 and
" Request for' Additional Information on Control of Heavy Loads".
To provide assurance that the evaluation criteria of NUREG-0612, Section 5.1'are met in addition to satisfying the general guide-lines of NUREG-0612, Section 5.1.1; a specific alternative identi-fled in NUREG-0612 has been selected for each evaluation of crane-design and operation in different areas.
A_ defense-in-depth approach is consistently used in the evaluation and control of heavy loads.
Several load drop analyses will be performed either to demonstrate compliance with the evaluation criteria or to determine any con--
sequent effects on the structure.
They are' scheduled to be completed by the end of June, 1982.
After all load drop analyses.are completed, it can then be determined whether any change or modification is required to fully satisfy the guidelines of NUREG-0612.
i 2
' Table 1-1 t
LOAD / IMPACT AREA MATRIX REACTOR CONTAINMENT BUILDING POLAR CRANE 4
i Tnentinn Reactor Containment Building
\\N Impact Azimuth'23 g& ~ Radius 63'-9" g
' Area Azimuth 337
& Radius 62'-9 Load Elevation Safety Related.
Hazard Elimin -
Equipment ation Category Main Hook Load above + 46' Vertical Duct-
. System Separa-Block (4. 5 tons) work tion i
+ 21' Containment fan System Eepara-coolers AH-1 tion.
(3A-SA) +
(3C-SA) i.
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Table 1-2 LOAD / IMPACT AREA MATRIX
' REACTOR CONTAINMENT-BUILDING POLAR CRANE LOCATION REACTOR CONTAINMENT BUILDING IMPACT Azimuth 148.5
+& Radius 42' +
7 AREA Azimuth 322.5
+& Radius 32.5 i LOAD-Elevation Safety Related Hazard.Elim-Equipment ination Cate-gory Main Hook Load
+46' Area' radiation System redun-Block (4.5 tons) monitors RE-24 dancy_and or any other heavy
& RE-25 separation loads
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