ML20076L125
| ML20076L125 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 07/13/1983 |
| From: | Harrington W BOSTON EDISON 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 83-181, NUDOCS 8307190016 | |
| Download: ML20076L125 (49) | |
Text
i BOSTON EntsDN COMPANY 800 BOYLsTON STREICT BOSTON M AssAcHuarTTs 02199
)
WILLIAM D. HARRENGTON sasesom vees passeosset
==*" July 15, 1983 BECo Letter No.83-181 Mr. Darrell G. Eisenhut, Director Division of Licensing ,
Office of Nuclear Reactor Regulation l U.S. Nuclear Regulatory Commission Washington, D.C. 20555 License No. DPR-35 Docket No. 50-293 Sebject: NUREG 0612: Control of Heavy Loads
References:
(A) NRC Letter, " Control of Heavy Loads," dated December 22, 1980.
(B) Boston Edison Letter from Mr. W.D. Harrington to Mr. Darrell G.
Eisenhut dated February 28, 1983.
Dear Sir:
In the letter of February 28, 1983, Boston Edison committed to provide a report which would satisfy the six and nine month reports as described in your letter of December 22, 1980.
Accompanying this letter is a report consicting of tv,0 enclosures. Enclosure 1 provides responses to Franklin Research Center's draft TER-C5257-109, dated February 12, 1982. Enclosure 2 responds to requests for information contained in Enclosure 3, Sections 2.2 and 2.3 of your December 22, 1980 letter.
We believe this report deconstrates that most of Pilgrim's lifting devices and procedures satisfactorily meet the guidelines of NUREG-0612. However, a number of procedural changes and modifications have been indicated necessary to ensure complete compliance with the goals of the NUREG. We shall complete all such modifications and procedure emplacements by December 31, 1984.
We believe this report satisfies the six and nine month report requirements.
Should you require any information regarding this submittal, please contact us.
Very truly yours, N
PMK/ mat Attachments: Heavy Loads Report Ao33 s 9307190016 830713 PDR ADOCK 05000293 P _PDR l lh j
- _d
ENCLOSUREI RESPONSES TO FRAWLIN RESEARCH CENTER DRAFT TER - CS257-109 DATED FEBRUARY 12,1982 EC Requests (from draf t Franklin TER Section 2.1.1.c)
Although FRC ogrees with the Licensee about the opplicability of NUREG-0612 to the reactor building bridge crone; insufficient information has been provided to allow on evaluation of compliance of other load handling systems with the guidelines. The Licensee should provide documentation of compliance or justification for exclusion.
RESPONSE
As a result of this FRC item and a better understanding of the criteria for excluding handling systems, BECo has reevaluated its initial response. Plant arrangement drawings, vendor equipment lists and area surveys were utilized to identify all handling systems that could carry heavy loads and to develop justification for the exclusion of porticular handling systems from the scope of NUREG-0612. The handling systems identified as being within the scope of NUREG-0612 ore:
Handling System Location Copacity (1) Reactor Building RB - l17' el. 100 ton (M)
Bridge Crone 5 ton (A)
(2) Turbine Building Bridge Turbine Building - 165 ton (M)
Crone SI' el. 25 ton (A)
(3) RHR Pump and Motor SE & NW Quadrants 5 ton /eo Hoists / Monorails (2) RB - 23' el.
(4) Recirculation Pump Drywell 20 ton and Motor Hoist /
Monorail (5) Fuel Pool & Reactor SE Quadrant 5 ton
{
Water Cleanup Filter RB - 91' el.
Equipment Hatch Hoists /Monorolls (2)
(6) Reactor Auxillory Boy Reactor Auxiliary 5 ton Equipment Hatch Holst/ Boy - 23' el.
Monorail (7) Recirculation Pump RB - 51' el. 8 ton MG Set Holst/ Monorail l
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A listing of the handling systems excluded and the basis for exclusion are provided below:
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a) Channel Hondling Boom This is a 200 lb. crane located on the refueling floor. The 200 lb. rating is less than that of a heavy lood, where o heavy lood is greater than the combined weight of a single spent fuel assembly and its handling tool (1500 lbs. for Pilgrim Station). No lood drop could result in domoge to ;
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safe shutdown or decay heat removal equipment, and this hoist cannot carry loods over the spent fuel pool or reactor vessel. Therefore, NUREG-0612 does not opply to ,
this equipment.
b) A & B Refuelino Jib Crones These are h ton cranes over the spent fuel pool on the refueling floor. One also may be used over the reactor vessel to old in refueling operations. The h ton roting is less than that of a heavy lood. Also, these crones do not travel over safe shutdown equipment. Therefore, NUREG-0612 does not opply to this equipment.
c) CRD Maintenonce Hoist / Monorail The 2 ton hoist for the CRD maintenance hoist is located in the CRD Maintenance Shop at elevation 23' in the southeast quadront of the Reactor Building. This holst/monoroll system is no longer in use therefore ,
NUREG-0612 requirements do not opply to this equipment.
d) Decontomination Room Holst This hoist is located near the North wall on elevation 23' of the reactor building. It is no longer used. Therefore, the requirements of NUREG-0612 do not apply to this equipment.
e) Fuel Rod Storoge Hoist The 2 ton fuel rod storage hoist is used to move fuel in the spent fuel pool and reactor vessel. This hoist is used only to move fuel and smaller tools i.e., no heavy loods
! ore handled, in addition, no safe shutdown or decoy heat removal equipment could be domoged by this crone.
Therefore, NUREG-0612 requirements do not apply to this ;
Crone.
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I Feedwater Heater A-Frome f) Reoctor Feed Pump Hoist These handling systems are located on elevation SI' of the l Turbine Building. Although there is cabling associated I with safe shutdown systems located at elevations below the 51' el. In this oreo, loss of these cables would not result in on inability to occomplish safe shutdown.
Therefore, NUREG-0612 requirements do not apply to this equipment.
g) Turbine Basement Hoist Rodsoste Area Bridge Crane Reactor Cleon-up Sludge Disposal Bridge Crone Off-gas Filter / Shipping Cask Monorail Hoist Retention Building Prefilter Monorail Hoist >
These crones are located in the rodwoste area of the plant, the main stock, or the Retention Building. No spent fuel or equipment necessary for safe shutdown or decoy heat removal is located in these areas. Therefore, the requirements of NUREG-0612 do not opply to this equi,) ment.
b) Electric Wire Rope Hoist Machine Shop Decontamination Trough Davit These cranes are located in the machine shop area of the plant. There is no spent fuel, safe shutdown or decay heat removal equipment that could be demoged by a load drop from either of these cranes. Therefore, the requirements of NUREG-0612 do not apply to this equipment.
- 1) General Area Monorail This 1500 lb monorail is located at the 91'3" el. of the Reactor Bui ding and troverses over the General Tool Coge, Receiving Coge, Relief Volve Cage and the Riggir g Tool Coge. The 1500 lb rating for this monorail system is less than that of a heavy lood for PNPS. Therefore, the requirements of NUREG-0612 do not apply.
j) Turbine Building Mezzonine Monorail This monorail is located at the 64' el. of the Turbine Building. It is used for maintenance of the Reactor and Turbine Building exhaust fans. A lood drop from this 3 I
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l monorail could not impoct any safe shutdown equipment. i Therefore, the requirements of NUREG-0612 do not apply ,
to this monorail, k) Diesel Generator Monoroils (4)
Eoch diesel generator has a monorail along each side running the entire length of the diesel generator. The monorails are only used for maintenance of their respective diesel generators. A drop from o monorail could only offect one diesel generator which would already be out of service for maintenance. Therefore, the requirements of NUREG-0612 do not apply to this monorail.
Tobles I, 2, and 3 provide listings of foods, load weights and load handling procedures for the Reactor Building Crane, Turbine Building Crane, and Monorail / hoist systems.
Evoluotion of Turbine Building Crone Against NUREG-0612, Section 5.1.1 -
General Guidelines The Turbine Building Crane was not previously oddressed in Boston Edison's June 25,1981 submittal. Information is provided below to address the General Guidelines in NUREG-0612, Section 5.1.1.
NUREG Section 5.1.l(l)- Sofe Load Poths As indicated in Table 2, there are two distinct load handling creas in terms of potential lood drop consequences associated with the Turbine Building Crone.
These are Load impact Regions 16 and 17 (see Figures 10 and 11 in Enclosure 2 to this submittol). Load drops in Region 16 will not offect the ability of safe shutdown systems to perform their safety functions.
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On this basis, and conservative structural, onolyses of the Region 17 floor area, lood bondling procedures will impose two levels of administrative control. As indicated in Enclosure 2, more rigorous structural analyses are being performed that may result in refinement of the specifics of these administrative restrictions. The first level of odministrative control is illustrated in Figure I and involves defining on exclusion area over which only certain heavy loods con be carried. All other heavy loods wil'l be handled by the Turbine Building Crane in Region 16 outside of this exclusion arco.
The second level of odministrative control involves defining specific safe load paths for certain limiting loads within Region 17. The loads involved are listed in Table 2. The safe food paths are illustrated in Figure 2. In addition to safe load paths, procedures will specify limits on carry heights within this region.
NUREG Section 5.I.l(2)- Procedures Lood handling procedures containing the information described in NUREG-0612, Section 5.l.l(2) will be developed and implemented for the Turbine Building Crane prior to movement of any heavy loads within the exclusion oreo.
Deviations from safe food paths will be controlled in the manner described in the response to draft TER Open item 2.1.2.c (poge 10 of this enclosure).
NUREG Section 5.1.l(3)- Crone Operators Crane operators will be qualified in accordance with ANSI B30.2.
NUREG Section 5.1.l(4)- Special Lifting Devices Special lif ting devices utilized with the Turbine Building Crone are used for lif ts over Region 16 only. Because load drop consequences in Region 16 will not offect the ability of safe shutdown systems to perform their sofety functions, comparison of these lif ting devices to the requirements of ANSI N14.6-1978 was judged not to be worronted.
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I NUREG-0612, Section 5.l.l(5) - Non Special Lif tino Devices Sling selection, use and maintenance will be controlled as indicated in the response to draf t TER Open item 2.l.6.b (page 19 of this enclosure).
NUREG Section 5.l.l(6)- Handling System Design The Turbine Building Crone was designed to crone industry stonderd EOCl-61. As a result, the Turbine Building Crane design has been evoluoted by verifying that the 10 CMAA-70 requirements identified in draft TER Open item 2.1.8.c have been satisfied or compliance justified by equivalent means. The results and conclusions of this comparison are provided in the response to drof t TER Open item 2.l.8.c (poge 20 of this enclosureh NUREG Section 5.1.l(7)- Handling System inspection, Testing and Maintenance Maintenance procedures will be implemented for the Turbine Building Crane that are consistent with the inspection, testing and maintenance guidelines of ANSI B30.2-1976.
Evoluotion of Monorails /Holsts Against NUREG-06I2, Section 5.1.1 - General Guidelines The handling systems (listed as 3 through 7 above) were not previously addressed in Boston Edison's June 25, 1981 submittal. Information is provided below to address the General Guidelines in NUREG-0612, Section 5.1.1 for each of these monorail / hoist systems.
NUREG Section 5.l.l(l)- Safe Lood Poths Load paths for monorail systems are defined by the limits of the monorail track.
In one case, however, limitations on movement along the monorail trock were judged to be necessory.
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Heavy load movements with the Recirculation Pump MG Set Monorail / Hoist are not generally anticipated during power operation. This is because Technical Specifications limit the time that a Recirculation System loop con be out of service to only 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to proceeding to the cold shutdown condition.
If heavy loods are ever moved in this area at power, then the concern becomes the potential for domoge to CRD hydraulic control units below the 51' el. deck in this oreo. On the basis of structural onalysis of load impocts on this deck,it was determined that certain odministrative controls were prudent for this ceto if heavy loads are to be moved at power. These odministrative controls are illustrated and described in Figure 3. They will be incorporated into load handling procedures applicable to this handling system.
NUREG Section 5.1.l(2)- Procedures Load handling procedures containing the information described in NUREG-0612, Section 5.l.l(2) will be developed and implemented for each of the identified monorail / hoist systems prior to lifting any heavy loads with these handling systems.
NUREG Section 5.1.l(3)- Crone Operators Hoist operators for the identified monorail / hoist systems will be qualified in accordance with ANSI B30.2-1976.
NUREG Section 5.l.l(4) - Special Lif ting Devices No special lifting devices have been identified that are used with the identified monorail /holst systems.
NUREG Section 5.l.l(5)- Non Special Lif ting Devices Sling selection, use and maintenance will be controlled as indicated in the response to draft TER Open item 2.l.6.b (page 19 of this enclosure).
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NUREG Section 5.1.l(6)- Handlino System Design The design of the following handling systems has been evoluoted as described below.
o RHR Hoist / Monorails (2) - 5 ton capacity o Recirculation Pump Motor Holst/ Monorail - 20 ton o Fuel Pool and Reactor Water Cleanup Filter Equipment Hatch Holst/ Monorails (2)- 5 ton o Reactor Auxiliary Boy Equipment Hatch Hoist / Monorail - 5 ton o Recirculation Pump MG Set Hoist / Monorail - 8 ton These handling systems are single monorail tracks suspended from Reactor Building, Reactor Auxiliary Boy, or Drywell structural I-beams. The criterio of ANSI B30.2 and CMAA-70 are not applicable to the design of handling systems such as these monorails and hoists. Accordingly the design of the monorail / hoist systems was compared to the criteria in opplicable standards, i.e. ANSI B30.II,
" Monorail Systems and Underhung Crones - 1980," and ANSI B30.16, " Overhead Hoists -1973."
Based on a point-by-point comparison to these standards, it was found that these monorail systems conform to the criterio in these current stondords, including requirements for maximum monorail deflection and monorail stress design safety factors with the exception of demonstrating a design safety factor of 4 for hoist components, and providing a warning label on hoists.
ANSI B30.16 requires that for hoist components the stress due to the rated load shall not exceed 25% of the overage ultimate material strength. Stress design safety factors are not available for the chain type hoists, and the complexity of these devices does not lend them to performance of stress analyses. However, safety margins were odequately oddressed in the purchase specifications by requiring rigorous load testing. The following summarizes the lood tests required l
in the purchase specifications for these hoists: l l
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(l) Proof testing of the book to 200% of rated capacity, followed by NDE to demonstrate no permanent deformation; (2) Proof test of the lood chain to 300% rated capacity; (3) Lood test by the supplier at 150% rated capacity for the completed hoist and trolley, prior to shipment; and (4) Performance of a 150% lood test of the hoists, trolley and monorail through raising, lowering and travelling operations. This test was performed offer final installation.
Additionally, inspection and maintenance procedures will be implemented that assure monorail and hoist components remain in good working condition. The load tests performed on these hoists and the continuing inspection and maintenance demonstrate on adequate achieved safety margin.
ANSI B30.I6 requires a WARNING label on the hoist or load block to caution personnel ogainst overloading; the danger of using twisted; kinked or domoged hoist cable, slings or hoist chains; using a demoged or malfunctioning hoist; lifting people; or operating the hoist with other than monval power. The Bechtel specification for these hoists did not require application of such a label, and on inspection of these hoists has verified that such warning labels are not offixed to these hoists. Boston Edison will add warning labels to these hoists that meet ANSI B30.16.
NUREG Section 5.1.l(7)- Handling System Inspection, Testing and Maintenance Maintenance procedures will be implemented for the identified monorail / hoist systems that are consistent with the inspection, testing and maintenance requirements of ANSI B30.ll-1980 and ANSI B30.16-1973.
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MtC Requests (from draft TER Section 2.1.2.c)
The Pilgrim Nuclear Power Station does not comply with Guideline I of NUREG-0612. To comply, the Licensee should perform the fallowing:
- 1. Verify that safe load paths are clearly marked.
- 2. Verify that deviations from established load pathways require written alternati,ves which must be specifically opproved by the plant safety review committee.
RESPONSE
- 1. Safe load path merkings - As indicated in Boston Edison's submittal of June 25, 1981, marking of safe lood paths on the refueling floor or the Turbine Building operating deck is not practical because of the extensive use of temporary coverings on the floor during outages. Further, current procedures and practices provide alternative visual aids to the crone operator during heavy load movements that meet the intent of this NUREG-0612 guideline. These practices are summarized below:
(a) The crone operator /signolman will verify the load path prior to load movement to assure that it is clear of obstructions.
(b) The signalmon will have load movement procedures including figures indicating the safe lood path in his possession or will have reviewed the specific load paths involved prior to movement and will direct the crone operator along the designated load path in occordance with the procedures.
- 2. Deviations from Safe Load Poths - Heavy food handling procedures which include safe load paths are safety-related procedures and accordingly changes to these procedures (including deviations from safe lood paths) are controlled. For PNPS this involves preparation of a safety evoluotion and review and opproval of the Operations Review Committee (ORC).
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NRC Request: (from & oft TER Section 2.1.S.b)
The Licensee has provided insufficient information to allow evoluotion of the compliance of special lifting devices of the Pilgrim plant with Guideline 4 of NUREG-0612.
RESPONSE
Special lif ting devices utilized with the Reactor Building Crone that have been evoluoted against the requirements of ANSI N14.6-1978 for the purpose of developing a response to this item are listed below:
- 1. Head Strongback
- 2. Dryer / Separator Lif ting Sling Assembly The dryer / separator sling assembly and the head strongback were evoluoted against ANSI N14.6 as described below.
Description of Dryer and Separator Sling The dryer and separator sling is used to remove and instcIl the dryer and the steam separator ossembly. The device is a cruciform steel frame ottoched to a hook box by four wire ropes with turnbuckles. The four ends of the cruciform frame are each fitted into o bell-shaped housing which is open and flored at the bottom. A hole posses through two sides of the housing for the lifting pin travel. Each lifting pin is octuated by a double-octing air piston. The lif ting pin, in turn, octuates on air volve at the end of the pin's travel. This air volve gives positive indication by way of a pressure gauge, that the lifting pin is fully inserted into the dryer and separator lifting lug. A lif ting eye, located on top of each I-beam, is connected to a turnbuckle and a wire rope. The wire ropes are ottoched to the hook box by spelter sockets and pins. The book box contains a
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slot at the top which is sized to accommodate the double hook II
of the crone. Two hook pins pass through the hook box to engoge the crone book.
Description of the Head Strongbock The heod strongbock is used to hoist the drywell head and the reactor vessel head. The device consists of four lifting arms mounted at right angles between top and bottom four-point stor p!ates. The top plate has o slot through which the double hook of the crane posses to engage the two hook pins. The strongbock is ottoched to lifting lugs on the drywell head and reactor vessel head, and to lifting lugs at the end of each arm of the strongbock, by turnbuckles and anchor shackles.
For the reasons listed below, the detailed comparison of the Dryer / Separator Sling Assembly and the Head Strongbock to ANSI N14.6-1978 was limited to Sections 3.2 and certain parts of Section S of the standard.
- 1) These devices were designed by General Electric Com-pony prior to the existence of ANSI N14.6-1978. In this regard, there are o number of sections in the standard that it is not reasonable to appiy in retrospect. These are the sections entitled, Designer's Responsibilities (Section 3.1); Design Considerations (Section 3.3);
Fabricator's Responsibilities (Section 4.1); inspector's Responsibilities (Section 4.2); and Fabrication Considerations (Section 4.3). Becouse documentation is not available to assure that all of the subparts of these sections were met, they have not been oddressed item by item for the purpose of identifying and justifying exceptions. However, information on the design drawings indicate that sound engineering proctices were placed on the fabricator and inspector by the designer for the 12 l
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purpose of ossuring that the designer's intent was occomplished. On this basis, there is reason-oble assurance that the intent of the sections of the stonderd listed above was, in fact, occomplished in the design, fabrication, inspection, and testing of these devices.
- 2) Section 1.0, Scope; Section 2.0, Definitions; Section 3.4, Design Considerations to Minimize Decontamination Ef-fects in Special Lif ting Device Use; Section 3.S, Cootings; Section 3.6, Lubricants; and Sections S.2.3, S.3.4 and 4.3.5 related to functional testing of non-load bearing ports are not pertinent to food handling reliability of the devices and, therefore, have not been oddressed for the purpose of identifying and justifying exceptions.
- 3) Section 6, Special Lifting Devices for Critical Loads, is applicable to critical loads. A critical lood is defined in the stonderd as:
"Any lif ted load whose uncontrolled movement or release could adversely offect any safety related system when such system is required for unit safety or could result in potential off-site exposures comparable to the guideline exposures outlined in Code of Federal Regulo-tions, Title 10, Port 100."
The applicability of Section 6.0 of ANSI N14.6-1978 is discussed in NUREG-0612 Section S.I.6. This Section of the NUREG indicates that special lifting devices utilized for heavy lood handling with crones that rely on upgrading to single-follure-proof to demonstrate complionce, should comply with ANSI NI4.6-1978, Section 6.0. The Reoctor Building Crane with which the two special lifting devices referred to obove are utilized has not been upgraded to single-failure-proof. Therefore, Section 6.0 does not apply.
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ANSI N14.6 - Section 3.2 Section 3.2 of ANSI N14.6-1978 establishes design criteria for special lifting devices. Specifically, it establishes (1) stress design factors for food bearing members and (2) brittle fracture criteria for materials used in lood-bearing members.
Design documents necessary to verify compliance or identify exceptions to these criteria are not available. Nonetheless, it is believed that adequate verification of the design safety margins have been demonstrated based on the following:
(1) Proof Load Tests - The Dryer and Separator Sling Assembly and Head Strongbock were required by drawing specification to be proof-tested at 125% of their rated capacity. Thorough visual, dimensional and NDE examinations were required following the proof test.
(2) In Service Examinations - Both devices have been utilized on many occasions to perform the lifts for which they were designed with no evidence of overstress or permanent deformation.
(3) The devices are utilized only for the lifts for which each was specifically designed. They have not been and will not be used for any other purpose. Therefore, the possibility of on overloed situation is extremely remote.
(4) As indicated in the discussion below regarding inspection and maintenance, the devices will periodically be subjected to necessary visual, dimensional, and nondestructive examination. This should assure that any indication of overstress will be detected and oction taken to repair or replace the domoged components.
ANSI N14.6 - Section 5 The comparison of current practices for inspection, testing, and maintenance of the dryer / separator sling and head strongback to Section 5 of ANSI N14.6-1978, I
. os supplemented by NUREG 0612, Section 5.l.I(4), found that certain changes to PNPS procedures were required in order to satisfy the inspection and test requirements in ANSI N14.6. These changes will be mode so that the PNPS inspection program complies fully, with the following exceptions:
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I Exception 1: Section 5.3.1 of the Standard requires that either
' lood testing or o comprehensive inspection program be undertaken annually for each special lifting device to assure continuing compliance with the standard. The option of periodic inspection has been chosen for the PNPS lifting
! devices. ,
ANSI N14.6 was developed to be opplicable "for special lif ting devices for shipping containers weighing 10,000 pounds or more for nuclear materials," most notably lifting devices for casks.
The service environment for lifting devices such as casks is different and generally more severe-than the service environ-ment for the PNPS lifting devices. Accordingly, it is our position that a less restrictive inspection program is worronted to assure continued serviceability for the PNPS lifting devices than that which is specified in ANSI N14.6. We propose that the full set of inspections prescribed in ANSI N14.6-1978, Section 5.3.l(2) be completed on a five-year interval.
Additionally, thorough visual examinations prior to each period of use of the lif ting devices will be undertaken. This inspection program is judged to be equivalent to the intent of ANSI N14.6
- ond to provide sufficient periodic inspection and examination to identify wear or degradation that could potentially reduce design safety margins.
The. bases for the extended frequencies for certain inspections are os follows:
- o. Frequency of Usage Since the lif ting devices identified for PNPS are typically used on on annual basis to support refueling operations, the frequency of use is considerably less than that of the
. special lifting devices for which ANSI N14.6 ' was developed. Special lifting devices for items such as casks 1 15
are potentially used between 50 to 100 times annually.
The reduced frequency of use limits the number of stress cycles to which the PNPS devices are subjected and, in turn, the cumulative usage factor and the potential for abuse and domoge.
o Controlled Environment The PNPS lifting devices are stored inside the Reactor Building in a dry, chemical free environment. All lif ting devices are inspected for cleanliness and cleaned prior to each use. On the contrary, the lif ting devices for items such as casks for which ANSI N14.6 was developed are subjected to harsh environments that may include rain, road dust, road salt, and other potentially deliterious materials, os well as greater abuse since they are transported on open truck flatbeds. Furthermore, os part of normal service, casks and their lifting devices must be decontaminated, which requires the use of various ocidic and caustic solutions. The obsence of potentially corrosive compounds and solutions lessens the likelihood of environmental service related domoge to the PNPS lif ting devices.
In conclusion, the service conditions are relatively mild and operating procedores provide substantial assurance that heavy loads will be handled in a safe manner, minimizing the potentiel for domoge of the lifting devices. The comprehensive PNPS l
visual and 5-year dimensional and nondestructive examination program will odequately confirm that design margins of safety l
have not been compromised due to potential service related mechanisms of degradotion.
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Exception 2: Plant procedures do not specify a visual inspec-tion by maintenance or other nonoperating personnel at inter-vols of three months or less os required by Section 5.3.7 of ANSI N14.6-1978. Between periods of usoge, these devices are stored in a specific location under o controlled environment and are not subjected to ony other usage except the dedicated and specific usoge mentioned in the description of the devices.
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Procedures require that the devices be inspected and examined by qualified personnel prior to usoge os described above. Based on the controlled storage, dedicated usage, and the complete inspection schedule, the equivalency of Section 5.3.7 is demonstrated.
Exception 3: Section 5.3.3 of ANSI N14.6-1978 requires that special lifting devices be lood tested according to Section 5.2.1 to 150% of maximum lood following any incident in which any food-bearing component may have been subjected to stresses substantially in excess of those for which it was qualified by previous testing, or following on incident that may have caused permanent distortion of lood-bearing parts. Since distortion may already have occurred or defects may have already de-veloped due to the overstressed condition, it seems more prudent and proctical to perform the dimensional examinations for deformation and the nondestructive examinations for de-fects to determine whether the device is still acceptable for use rather than to subject the device to 150% load testing. If major repairs are required, the device shall be repaired or modified and then tested to 150% lood followed by examination for defects or deformation. Major repairs are defined in Section 5.3.2 of ANSI NI4.6-1978. This alternative ochieves the
! some objective os Section 5.3.3 of the standard.
Exception 4: Section 5.2.1 of the_ standard requires on initial load test of 'l50% of rated load. The PNPS special lifting 17
devices were subjected to proof lood tests of 125% of rated load. This is consistent with industry standards for other heavy lifting qquipment such as crones and therefore, is judged to be odequate to demonstrate lood carrying capacity substontially in excess of roted food.
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1 titC Requests (from & aft TER Section 2.l.6.b)
The Licensee hos provided insufficient information to allow evoluotion of i compliance of the Pilgrim plant's not-specifically-designed lifting devices with Guideline S of NUREG-0612.
RESPONSE
Heavy load handling procedures for all handling systems included within the scope of NUREG-0612 require that sling selection be based on the opplicable sections of ANSI B30.9-1971 for various sling types. This involves:
(1) An occurate designation of the food weight, (2) the addition of a dynamic load factor to the food weight of 1/2% of the lood weight for each it/ min of hoist-speed (this factor is designated in the procedures for specific hoists), and (3) selecting a sling of appropriate capacity for the combined food weight and dynamic loading factor.
All slings will be inspected, tested, repaired and replaced in occordance with the opplicable sections of ANSI B30.9-1971.
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NRC Request: (from drof t Freklin TER, Section 2.1.8.c)
The Pilgrim Nuclear Power Station complies with NUREG-0612, Section 5.1.1, Guideline 7, to a substantial degree, on the basis of compliance with EOCl-61 criteria. However, the Licensee should provide information to verify that (1) the following CMAA-70 requirements have been satisfied for cranes subject to this review or (2) the requirements of CMAA-70 have been satisfied by equivalent means:
- 1. nonsymmetrical girder sections were not used in construction of the cranes
- 2. any longitudinal stiffeners in use conform to the requirements of CMAA-70, and allowable h/t ratios in box girders using these stiffeners do not exceed ratios specified in CMAA-70
- 3. girders with b/c ratios in excess of 38 were not used
- 4. fatigue failure was considered in crane design and the number of design loading cycles at or near rated load was less than 20,000
- 5. the sum of maximum crone load weight and the weight of the bottom block, divided by the number of parts of rope, does not exceed 20% of the manufacturer's published rope breaking strength
- 6. drum desyn calculations were based on the combination of crushing and bending loods
- 7. drum groove depth and pitch conform to the recommendations of CMAA-70
- 8. mechanical lood brakes or hoist holding brakes with torque rotings of approximately 125% of the hoist motor torque were used
- 9. ony static control systems in use conform to the requirements of CMAA-70
- 10. controllers used were of the spring-return or momentary-contact pushbutton type.
Response
The requested verification is provided below for the Reactor Building and Turbine Building Overhead Bridge Cranes.
Reactor Building Crane The PNPS Reoctor Building Crone was built prior to the issuance of ANSI B30.2-1976 and CMAA 70-1975. This crone was designed and fabricated by Crone 20
l l
1 Corporation in accordance with EOCl-61, l Manufocturing and Service
" Specifications for Electric Overhead Traveling Crones-1961," and additional criterio contained in Bechtel Specification No. 6498-M-23, Rev. I, October 24, 1968. These specifications addressed certain, but not all, of the criteria in ANSI B30.2-1976 ond CMAA 70-1975. To oddress the 10 points identified in the Franklin Research Institute's TER where CMAA-70 and ANSI B30.2 ore more restrictive than EOCI-61, o design evoluotion of the PNPS Reactor Building Crone was performed. The following summarizes our findings for these 10 points.
(1) Torsional Forces - CMAA 70 specifies that twisting moments be determined based on the horizontal distance between the center of gravity and the shear center of the girder section. EOCI-61 requires twisting moments to be based on the distance between the load center of gravity ond the beam center of gravity. Since the PNPS Reactor Building Crone girders are symmetrical box sections, these two requirements are the same. Since the trolley rails are located over the centerline of the girders, there are no opprecioble torsional forces on the girders. Thus PNPS Reactor Building Crone satisfies CMAA 70 criterio relative to torsional forces.
(2) Longitudinal Stiffeners - CMAA 70 specifies o minimum moment of inertio for longitudinal stiffeners, maximum width to thickness ratio, and stiffener location along the web plate. EOCl does not provide similar guidance. For the PNPS Reactor Building Crane, application of the CMAA criterio requires that the moment of inertio be greater than Ic = 52.15-in.4, the width to thickness ratio should be less fl on 12, and the stiffener should be located
' O.4 of the distance from the compression plate to the web neutral oxis. The octual moment of inertio is 159-in.4, the stiffener width to thickness ratio is 9.6, and the stiffener centerline is located 0.43 of the distance from the 9
21 i
t
)
compression plate to the web neutral oxis. Thus the CMAA criteria relative to longitudinal stiffeners are set'sfied for this crone.
CMAA 70 specifies that 1/h (I = girder spon; h = web
. height) should be less than 25; EOCl-61 has no limit on I/h.
For the PNPS Reactor Building Crone, I/h = 1228 in./84 in.
= 14.6. Therefore, CMAA 70 is satisfied.
In addition, CMAA 70 specifies that h/t be less than C(K+1) 17.6 and less than M, where:
fc t = web thickness = 3/8 in.
C = 162 (the PNPS Reactor Building Crane has one
~
longitudinal stiffener)
K = f t/fc = 1.0 ft = max. tensile stress = 16.0 ksi fe = max. compressive stress = 16.0 ksi M = 376 Therefore occording to CMAA 70, h/t should be less than 340.2 and less than 376. h/t = 84/(3/8) = 224. Therefore, CMAA 70 is satisfied.
(3) Basic Allowable Stresses - EOCl-61 is more conservative than CMAA 70 for allowable tension, compression, and shear stresses, if b/c is less than 38 (b is distance between web plates and e is the thickness of the cover plate). For the PNPS Reactor Building Crane, b/c is 24 in./l.5 in. =
- 16. Therefore, CMAA 70 is satisfied.
22 l
(4) Fatigue Failure and Cyclic Loading - CMAA 70 specifies that fatigue failure be considered in the crane design, and also specifies on allowable stress range for crone structural members that are subject to cyclic looding of greater than 20,000 over the life of the crone. The number of cycles for any, of the crone members will be less than 2,000 over the life of the PNPS Reactor Building Crone. Based on this, failure due to cyclic fatigue should not be of concern for this crone, and the CMAA 70 criteria for cyclic loading are satisfied.
(5) Hoisting Rope - CMAA 70 specifies a 5:1 hoisting rope safety foetor for the rated load plus bottom block divided by the number of ports of rope. For the PNPS Reactor Building Crane, the resulting safety factor for the main hoist is:
bottom block = 6,100 lbs.
rated load = 200,000 lbs.
ports of rope = 12 (1-1/8" each) rope published breaking strength = 113,000 lbs.
resulting safety factor = 113,000/((200,000+6,100)/12)=6.57:1 For the aux. hoist:
block = 20 lbs. (81/2 ton capacity) rated lood = 10,000 lbs.
parts of rope = I (7/8" dia. - Type 304) rope published breaking strength = (g.t.) 56,000 lbs.
resulting safety factor = 56,000/10,020 = 5.6:1 Therefore the ropes satisfy the criteria in CMAA 70.
23
(6) Hoist Drum Loods - CMAA 70 specifies that drum design should consider combined crushing and bending loods; however, EOCl 61 is not as specific. The Bechtel design specification for this crane required the design to consider combined crushing and bending loads. Therefore CMAA 70 is satisfied. ,
(7) Hoist Drum Groove - CMAA 70 specifies minimum drum groove depth and drum groove pitch; EOCl 61 does not provide such specific guidance. For the Pilgrim Reactor Building Crane, this guidance would require minimum drum groove depth and pitch of 0.42 in. and 1.25 in, respectively for the main hoist, and 0.33 in. and 1.0 in.
for the oux. hoist. The octual dimensions are 0.4375 in.
and 1.25 in, for the main hoist and 0.344 in. and 1.0 in, for the oux. hoist. Thus, the CMAA criteria are satisfied.
(8) Hoist Holding Brakes - CMAA 70 and ANSI B30.2 require that holding brakes have minimum torque ratings (relative to motor terque) of 125% if used with control broking other than mechanical; 100% if used with mechanical control broking.
For the PNPS Reactor Building Crane, the design specification called for brakes that are 150% of the full rated torque of the hoist motor, for both the main and auxiliary hoists. The main hoist brakes octually installed have o total torque rating of 370% of the full rated motor torque. The hoist holding brakes are, however, provided with a time delay so that both brakes are not applied at the some time. Each broke thus has o rating of 185% full
, motor torque.
F l
l The auxiliary hoist also has two holding brakes, applied with a time delay. Eoch broke has a torque rating of l 200% full motor torque.
1 Therefore the holding brakes satisfy CMAA 70.
(9) Static Controls - CMAA 70 includes various criterio for crone static controls; EOCl only addresses crone magnetic controls. Since the Pilgrim Reactor Building Crone uses a D.C. mognetic control system, the CMAA 70 criterio on static controls are not applicable.
(10) Restort Protection - CMAA 70 establishes criteria for restort protection for crones not provided with spring-return controllers or momentary contact pushbuttons; this is not oddressed in EOCl 61. These CMAA 70 criterio are not applicable to the PNPSReactor Building Crane since this crone has spring-return pushbutton controls.
Turbine Building Bridge Crone The PNPS Turbine Building Bridge Crone was also built prior to the issuance of ANSI B30.2-1976 and CMAA 70-1975. This crone was designed and fabricated by Whiting Corporation in occordance with EOCI-61, " Specifications for Electric Overhead Traveling Cranes-1961," and additional criteria contained in Bechtel Specification No. 6498-M-12, Rev. I, February 19, 1968. These specifications addressed certain, but not all, of the criteria in ANSI B30.2-1976 ond CMAA 70-i 1975. To oddress the 10 points identified in the Franklin Research Institute's TER where CMAA-70 and ANSI B30.2 are more restrictive than EOCl-61, o design evoluotion of the PNPS Turbine Building Crone was performed. The following summarizes our findings for these 10 points.
(1) Torsior.ol Forces - CMAA 70 specifies that twisting moments be determined based on the horizontal distance 25
_ - - - - - , , v_ _ q --
between the center of gravity cnd the shear center of the 91 rder section. .EOCl-61 requires twisting moments to be l
based on the distance between the food center of gravity and the beam center of gravity. Since the PNPS Turbine
' Building Crone girders are symmetrical box sections, these two requirements ore the same. Since the trolley
' rails are located over the centerline of the girders, there are no oppreciable torsional forces on the girders. Thus the PNPS Turbine Building Crane satisfies the CMAA 70 criterio relative to torsional forces.
(2) Longitudinoi Stiffeners - CMAA 70 specifies a minimum moment of inertio for longitudinal stiffeners, maximum width to thickness ratio, and stiffener location clong the web plate. EOCl does not provide similar guidance. For the PNPS Turbine Building Crone, application of the CMAA criteria requires that the moment of inertio be
. greater than lo = 169-in.4, the width to thickness ratio should be less than 12, and the stiffener should be located 0.4 of the distance from the compression plate to the web neutral oxis. The octual moment of inertio is Sf 3-in.4, the stiffener width to thickness ratio is 11, and the stiffener centerline is located 0.S6 of the distance from the compression plate to the web neutral oxis. The CMAA criteria on stiffener moment of inertia and width to thickness ratio are satisfied. Location of the longitudinal stiffener closer to the neutral oxis means that less resistance to buckling of the web plate is provided, for the some stiffness afforded by the stiffener. Since for this crone, the octual stiffener moment of inertia is significantly above .the CMAA 70 minimum (513/169 =
greater than 300%), the design of the stiffener
- compensates for the deviation from CMAA 70 in the location of the longitudinal stiffener, and provides on i
odequate alternative to the CMAA criterio.
26
, .-e. .e 6-. er r , ,e -
.4..c.. - ,., ev%- e + ,,-,-m4 ..=, , , - -,m, ,- ,m.. _ . ,
i CMAA 70 specifies that 1/h (l = girder spon; h = web height) should be less than 25; EOCl-61 has no limit on 1/h.
For the PNPS Turbine Building Crane, I/h = 122d in./84 in.
= 14.6. Therefore, CMAA 70 is satisfied.
In addition, CMAA 70 specifies that h/t be less than I
17.6 and less than M, where:
C(K+1) t = web thickness = 3/8 in.
C = 162 (the PNPS Turbine Building Crane has one
!- longitudinal stiffener)
K = f t/fc = 1.0 ft = max. tensile stress = 16.0 ksi fe = max. compressive stress = 16.0 ksi M = 376 Therefore according to CMAA 70, h/t should be less than 340.2 and less than 376. h/t = 84/(3/8) = 224. Therefore, CMAA 70 is satisfied.
(3) Basic Allowable Stresses - EOCl-61 is more conservative than CMAA 70 for allowable tension, compression, and shear stresses, if b/c is less than 38 (b is distance between web plates and e is the thickness of the cover plate). For the PNPS Turbine Building Crane, b/c is 19.75 in./l.875 in.
= 10.5. Therefore, CMAA 70 is satisfied.
(4) Fotique Failure and Cyclic Loading - CMAA 70 specifies that fatigue failure be considered in the crone design, and also specifies on allowable stress range for crone
!- structural members that are subject to cyclic loading of greater than 20,000 over the life of the crane. The i
I 27
number of cycles for any of the crone members will be less than 13,200 cver the life of the PNPS Turbine Building Crone. Based on this, failure due to cyclic fatigue should not be of concern for this crone, or.d the CMAA 70 criterio for cyclic loading are satisfied.
(5) Hoisting Rope - CMAA 70 specifies a 5:1 hoisting rope safety factor for the rated lood plus bottom block divided by the number of ports of rope. For the PNPS Turbine Building Crane, the resulting safety factor for the main hoist is:
bottom block = 18,000 lbs.
rated lood = 330,000 lbs.
ports of rope = 16 (ik" each) rope published breaking strength = 123,000 lbs.
resulting safety factor = 123,000/((330,000 + 18,000)/16)=5.66:1 For the aux. hoist:
block =
rated load = 50,000 lbs.
ports of rope = 12 (9/16" each) rope published breaking strength = 26,000 lbs.
resulting safety factor = 26,000/((50,000+1800)/12)=6:1 Therefore the ropes satisfy the criteria in CMAA 70.
(6) Hoist Drum Loads - CMAA 70 specifies that drum design should consider combined crushing and bending loods; however, EOCI 61 is nc,t as specific.
A Whiting Corporation representative stated that the drum design L
for Whiting crones considers both crushing and bending loads. Therefore, CMAA 70 is satisfied.
l 28 I
- - - , - - - ._x-- g . -u ,t-2 ,, w,-_= .
1 (7) Holst Drum Groove - CMAA 70 specifies minimum drum groove depth and drum groove pitch; EOCl 61 does not provide such specific guidance. For the PNPS Turbine Building Crane, this guidance would require minimum drum groove ' depth and pitch of 0.47 in and 1.375 in.
respectively for the main hoist, and 0.21 in. and 0.69 in.
for the aux. hoist. The octual dimensions are 0.47 in, and
' l.375 in. for the main hoist and 0.21 in. and 0.69 in, for the aux. hoist. Thus, the CMAA criteria are satisfied.
(8) Hoist Holding Brokes - CMAA 70 and ANSI B30.2 require that holding brokes have minimum torque rotings (relative to motor torque) of 125% if used with control braking other than mechanical; 100% if used with mechanical control broking.
For the PNPS Turbine Building Crone, the design specification coiled for brakes that are 150% of the full rated torque of the hoist motor, for both the main and auxiliary hoists. The main hoist broke octually installed has a total torque rating of 168% of the full rated motor torque.
The auxiliary hoist holding broke has a torque rating of 230% full motor torque.
i Therefore the holding brakes satisfy CMAA 70.
(9) Static Controls - CMAA 70 includes various criteria for crone static controls; EOCl only oddresses crane magnetic controls. Since the PNPS Turbine Building Crane uses o i
mognetic control' system, the CMAA 70 criteria on static
- controls are not opplicable, 29 f
(10) Restort Protection - CMAA 70 establishes criteria for restart protection for crones not provided with spring-return controllers or momentary contact pushbuttons; this is not oddressed in EOCl 61. The PNPS Turbine Building Crone motion control is provided in the cab by GE General-Duti Surface-Mounted Master Switches which do not have spring return. The control circuitry design includes on undervoltoge release which prevents motor restort unless these controllers are in the neutral position.
Crone motion control from the floor is provided by a pendant-mounted pushbutton station. These push buttons are spring-return. The CMAA 70 criteria are satisfied.
30' l l
l
MtC Request (from & oft TER Section 2.2.1)
Technical Specifications (Interim Protection Measure I, NUREG-0612, Section 5.3(l))
" Licenses for all operating reuctors not having a single-failure-proof overhead crone in the fuel storoge pool area should be revised to include o specification comparable to Standard Technical Specification 3.9.7,' Crone Travel - Spent Fuel Storage Pool Building,' for PWR's and Standard Technical Specification 3.9.6.2,
' Crone Travel,' for BWR's, to prohibit handling of heavy loads over fuel in the storoge pool until imrlementation of measures which satisfy the guidelines of Section S.I."
- o. Summary of Licercee Statements and Conclusions The Licensee made no statement and expressed no conclusions regarding this interim protection measure.
RESPONSE
PNPS procedures define specific safe load paths for heavy loods handled in the vicinity of the spent fuel pool. These safe lood paths assure that no heavy loads l
pass over the spent fuel pool with the exception of 2 loads which must pass over certain peripheral areas of the pool. The 2 loads are casks which are corried over the cask loydown crea and the spent fuel gate, which is moved within the pool along the wall odjacent to the reactor cavity and periodico!!y lifted out of the pool for maintenance. None of these safe food paths pass directly over spent fuel in the spent fuel pool. In addition, at certain times the Reoctor Building i Crane load block may be moved over the spent fuel pool. The load block has not been postulated to drop into the spent fuel pool on the basis that redundant upper limit switches will be provided and redundant holding brakes exist for the crane hoists. These redundant devices will ossure that the probability of a two-blocking or uncontrolled lowering event will be sufficiently small to eliminate consideration of a postulated food block drop.
The procedures that include the safe food paths are safety related procedures and therefore, deviations are tightly controlled as described previously in the response to the open item related to draft TER Section 2.l.2.c. No odditional procedural controls or technical specifications are judged to be necessary at this time.
31
tftC Request: (from draf t TER Section 2.2.3) f l Special Reviews for Heavy Loods Over the Core (Interim Protection Measure 6, NUREG-0612, Section 5.3(1)
"Special attention should be given to procedures, equipment, and personnel for the handling of heavy loods over the core, such as vessel internals or vessel inspection tools. This special review should include the following for these loods: (1) review of procedures for installation of rigging or lifting devices and movement of the lood to assure that sufficient detail is provided and that instructions are clear and concise; (2) visual inspections of lood bearing components of crones, slings, and special lif ting devices to identify flows or deficiencies that could lead to failure of the component; (3) oppropriate repair and replacement of defective components; and (4) verify that the crane operators have been properly trained and are familiar with specific procedures used in handling these loads, e.g., hand signals, conduct of operations, and content of procedures."
- o. Summary of Licensee Statements and Conclusions The Licensee mode no statements and expressed no conclusions regarding this interim protection measure.
RESPONSE
Items (2), (3) and (4) are odequately oddressed in previous responses in Boston Edison's submittal of June 25, 1981 as supplemented by responses in this submittal. The review of PNPS procedures requested in item (l) hos been performed and it hos been confirmed that sufficient detail and clarity are provided for the subjects of interest.
l 32 l
- p. - ww m
TABLEI HEAVY LOADS - REACTOR BUILDING CRANE Heavy Loods Appror. Weight Procedures (Tons)
Woste Debris 25 6.9-167 Shipping Cosks Vessel Head 10 3.M.4-48 Insulation l Cattle Chute S 3.M.4-48 Head Strongback 2 3.M.4-48 Reactor Shield Plug (9) (3) 68 3.M.4-48 (6) 72 Drywell Head 43 3.M.4-48 Reactor Vessel Head 81 3.M.4-48 Steam Dryer Assembly 27 3.M.4-48 Moisture Separator Assemly 42.5 3.M.4-48 Spent Fuel Pool Gates (2) 0.6 (Max) 3.M.4-48 Dryer / Separator Storage 43 3.M.4-48 Pit Shield Plugs Refueling Slot Plugs (4) 6.6 (Max) 3.M.4-48 Spent Fuel Shipping Casks 26 6.9-164 Vessel Service Plotform 4 3.M.4-48 Miscellaneous Tool Boxes 3 New Fuel Crates I
- General heavy loods handling procedure to be implemented
l l
TABLE 2 WAVY LOADS - TURBINE BUILDING CRAE l
Loads Lifted in Region 16 - Lood Drops in Region (6 Will Not Result in Loss of Safe Shutdown Capability - Lif ts of These Loods Will Be Excluded From Region 17 Heavy Loads Approx. Weight Procedures (Tons)
H.P. Turbine Outer Shell 72.7 H.P. Turbine Rotor 61.7 L.P. Turbine Hoods (2) 57.8 L.P. Turbine Inner, Upper 60.3 Casing (2)
L.P. Turbine Rotor (2) 147.2 Generator Outer Shield (Upper) 6.6 Generator Rotor 15 L.P. Rotor Lif ting Beam 6T Condensate Pumps Motor Il.6 Pump a.8 Alternotor Rotor 7.5 Stortor 1.5 Miscellaneous Vorles Turbine Ports, Smo!! HX's, Steam Volve Components, Tool Boxes, etc.
Loads Lif ted in Region 17 - Load Drops in Region 17 Potentially Have Safety l
Consequences - see Enclosure 2 for Evoluotion l Feed Water Heaters / Drain Coolers
! E103A&B 34.25
! E104A&B 26.6 E105A&B 22.2 l
E106A&B 20.5 l
TABLE 2 l-EAVY LOADS - TURBlNE BUILDING CRAE (Continued) 9 Heavy Loads Approx. Weight Procedures (Tons)
Reactor Feed Pump & Motor Rotor 8 (Max)
Stortor 8 (Max)
Pump 4 (Max)
L.P. Turbine Diophrogms 2.5-10 (44 pieces)
- - General heavy loods handling procedure to be implemented I
1
l TABLE 3 L
l-EAVY LOADS - HOIST /MONORAll SYSTEMS Approx.
Monorail / Hoist Heavy Loads Wat (Tons) Procedures RHR Pump and RHR Motor 2.35 Motor Hoists (2)
Reactor Recire Recirculation 14 Pump and Motor Pump Motor Hoist Fuel Pool and Hotch Cover 3/4 Reactor Water Clean Up Filter Hoists (2)
Reactor Auxiliary Hotch Cover 3/4 Boy Equipment Hatch Hoist Recirculation Pump Fluid Drive 1/2-I MG Set Monorail Pumps and Motors Fluid Drive Cooler 1.8 Motor Generator 4.5 Armature
- Load handling procedures will be developed by Boston Edison prior to any future heavy load l lif ts with these handling systems.
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Note I: The following heavy loads may be lif ted by the Turbine Building Crane within the exclusion area. However, lifI height and procedural restrictions must be followed when such lif ts are made.
(1) Turbine Diaphragms (3) Feedwater Drain Coolers: E106A&R (4) Reactor Feed Pump Motor and Pump Components (2) Feedwater Heaters: E103A&B, E105A&B cuid E104A&B FIGURE I EXCLUSION AREA FOR ALL WAVY LIFTS BY TK TURBlNE BUILDING CRAE EXCEPT THOSE LISTED IN NOTE I TURBINE BUILDING PLAN EL. SI'-0"
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ie FIGURE 2 INTERIM SAFE LOAD PATH FEEDWATER WATER AND REACTOR FEED PUMP COMPOENTS TURBINE BUILDING SI' EL.
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If heavy lif ts are to be mode during power
- operation, the centerline l of the hoist hook is not to ]
j be moved over the shoded l
oreos until the lood has been lowered to o height 1 foot or less above the g sleideck.
L i
FIGURE 3 EXCLUSION AREAS RECIRC. PUMP MG SET MONORAll EL. SI'-D'
ENCLOSURE 2 RESPONSES TO REQUESTS FOR lif0RMATION IN SECTIONS 2.2 AtO 2.3 OF titC LETTER DATED 12/22/80 OVERVIEW OF NUREG-06l2 EVALUATlONS Boston Edison has identified the following fixed handling systems at Pilgrim Nuclear Power Station (PNPS) to which NUREG-0612 is applicable. They are:
Handing System Location Capacity Reactor Building RB - 117' el. 100 ton (M)
(I)
Bridge Crone 5 ton (A)
(2) Turbine Building Bridge Turbine Building 165 ton (M)
Crane 51' el. 25 fon (A)
(3) RHR Pump and Motor SE & NW Ouodronts 5 ton /eo Hoists / Monorails (2) RB - 23' el.
(4) Recirculation Pump Drywell 20 ton and Motor Hoist /
Monoroll (5) Fuel Pool & Reactor SE Ouodront 5 ton Water Cleanup Filter RB - 91' el.
Equipment Hatch Hoists / Monorails (2)
(6) Reactor Auxiliary Boy Reactor Auxiliary 5 ton Equipment Hotch Hoist / Boy - 23' el.
Monorail (7) Recirculation Pump RB - 51' el. 8 ton MG Set Hoist / Monorail There are two besic approaches available to demonstrate compliance with the NUREG-0612 guidelines. They are: (1) demonstrate odequate load handling reliability, or (2) demonstrate that load drop consequences are within the limits of Criterio I-IV listed in Section 5.1 of the NUREG. In all cases for PNPS, the 4
opprooch has been a demonstration of occeptable consequences.
I
A combination of systems, structural, cr.iticality and dose evoluotions has been utilized to address the NUREG-0612 guidelines for PNPS. To assist in performing these evoluotions the relevant plant creas were subdivided into potential load impact regions. The Reactor Building was subdivided into fourteen (15) load impact regions; eight of the regions on the 117' el. refueling floor and the others corresponding to the areas over which the various monorails could travel. The Turbine Building was subdivided into two (2) load impact regions. The subdivisions were based, in port, on the configuration of the buildings and a knowledge of specific locations where heavy loads are typically handled. The food impact regions are described in Table I of this enclosure.
Major component loydown creas for the refueling floor are illustrated in Figure 1.
The eight lood impoet regions for the Reoctor Building refueling floor are illustrated in Figures 2 through 9. The two load impoet regions for the Turbine Building are illustrated in Figures 10 and 11.
Tables I, 2 and 3 in Enclosure I to this submittal identify the loads and opplicable lood handling procedures for the identified handling systems.
Table 2 of this enclosure relates the defined food impact regions with the four criteria listed in Section 5.1 of NUREG-0612. As indicated in the table, the majority of the regions were defined to assist with and focus safe shutdown evoluotions to address NUREG-0612 Criterion IV.
Toble 3 of this enclosure indicates the types and combinations of evoluotions employed by Boston Edison for each region.
Evoluotion Methodology Systems Evoluotions As noted above, systems evoluotions were utilized for many of the regions to evoluote potential lood drop consequences. The objective of these systems evoluotions was to determine whether defined safety functions could be occomplished assuming that certain equipment was made inoperable os a result l
of a postulated load drop.
l 2
\
1 I
The steps used to perform systems evoluotions of the potential effects of lood j
drops inside the reactor building are outlined below: l 1) define the safety functions which must be occomplished;
- 2) identify the systems and, support systems relied on to occomplish each safety function;
- 3) for each lood impact region, identify the opplicable safety functions and systems and determine which components of those systems could be offected by a heavy load drop within the region;
- 4) perform on analysis of the effects of failure of those components on the ability to accomplish the opplicable safety functions.
The safety functions and systems defined for the purpose of performing the The results and PNPS systems evoluotions are illustrated in Figures 10 and 11.
conclusions of the systems evoluotions are described in the responses to Request for Information item 2.3.
Structural Analyses As indicated in Table 5, a number of load impact regions were addressed utilizing structural analyses /evoluotions. The load drop scenarios addressed or supported with structural analyses included:
- 1) drops onto and into the reactor vessel;
- 2) drops onto the spent fuel pool floor concrete stab, and; l
, 3) drops onto concrete slobs over safe shutdown equipment. :
i e. -
3
The results and conclusions of the structural analyses are described as oppropriate in the responses to Request for Information items 2.2 and 2.3. The steps in the structural evoluotion opproach and general methodology are discussed below:
The general steps used to perform the structural evoluotions are outlined below:
- 1. Identification of heavy lood, handling systems, and hand-ling locations including a full chorocterization of the food weight, dimensions, material properties, and structural chorocteristics.
- 2. Development of postulated drop scenarios based upon realistic consideration of plant procedures.
- 3. Review of important structural engineering aspects of impacted structural elements to fully chorocterize behavior. For reinforced concrete and steel elements, identify drops which control "locol" response (e.g.,
penetration, scabbing, spalling, perforation, etc.); loods that control "overall" structural response (e.g., large inelostic deformations or obrupt failure of principal struc-tural members, etc.); ond/or loods that may induce behavior that exhibits combined response such that either overall or local failure modes would control.
- 4. Incorporating 1 through 3 obove, provide early input to the systems evolvations to foetor structural informatinn into systems evoluotions assumptions.
- 5. Conduct detailed structural evoluotions that include:
- a. Specification of impact energy considering, os appropriate, the energy dissipated due to the trans-fer of momentum, fluid drag, bouyancy, etc.;
- b. Model develoment for ossessing dynamic response !
I utilizing empirical dato os necessory;
- c. Development of failure criteria based upon stability or leak tightness considerations;
- d. Computation of the strain energy obsorbed prior to reaching the prescribed performance limits;
- e. Assessment of structural response and structural i
'* consequences of drop. .
4
The structural evoluotion methodology and criteria generally follow the recommendations made by the American Society of Civil Engineers Technical Committee on Impulse and impoet Loads (Reference 1). These recommendations are supplemented by a large body of experimental and analytical information which is documented in reports which have been published by government, university, and industry organizations..
The evoluotion methodology and criteria which are addressed below consider the j
{
two potential modes of structural behavior, local effects and overoll structural response, respectively.
Local Impact Response Evoluotion Local impact response may lead to severe domoge such as crushing, perforation, and concrete ejection in the vicinity of the impactive food; however, overall dynamic response of the structure in the form of reactions away from the food are insignificant. The complex nature of local impoet response of reinforced f
concrete requires evoluotion using empirical formulae that are experimentally derived. The modified National Defense Research Committee (NDRC) formula (Reference 2) was chosen because it has been shown to give the best fit with ovoilable experimental dato (References 3 and 4). The NDRC formulae for the depth of penetration, x (inches), of a solid cylindrical missile are given by:
1.8- 1/2 for 5 4 (1) x= 4KNWd V d ~ 2.0
. TODUd)/ .
or
~
-l*8 x=KNW V +d for E > 2.0 (2) 1000d, d -
l 1
where W = weight of the missile (pounds) i d = diameter of missile (inches)
- t. V = impoet velocity of missile (feet /second) ,.
N = missile shape factor 5
= 0.72 flat-nosed missiles
= 0.84 blunt-nosed missiles
= 1.00 spherical-nosed missiles
= 1.14 sharp-nosed missiles K = concrete penetrobility factor
= 180/,ff'c (f'c = concrete compressive strength in pounds / square inch)
The thickness of reinforced concrete needed to resist impact without perforation and scabbing are given by the following Army Corps of Engineers formulae which con be used in conjunction with equations I and 2 (Reference 5).
i (3) ts = 2.12 + l.36 [*\ for 0.65~d 4 5 4 Il.75
~
d )
f* for 1.35 4 5 4 13.5 (4) tg = 1.32 + 1.24 d (d
d (
where is = concrete thickness required to prevent scobbing t p = concrete thickness required to prevent perforation i
Equations 3 and 4 were later extrapolated for small values of x/d (Reference 6) giving, l
5 5 for 5 40.65 (5) ts = 7.9I - 5.06 j d (dj d d g, r3 t2 (6) 5 for 5 f. l.35 tg = 3.19 I 5 - 0.718 f (d) d d M A 10 percent margin on thickness has been opplied in the use of equations 3 through 6 os recommended in Reference I, except for concrete sections backed by steel decking where the equations were used directly.
6
The effects of shape and deformability have been conservatively occounted for in the cose of the NDRC formula by adjusting the missile shape foetor, N, and/or using " equivalent" diameters.
' Overall Structural Response Evoluotion i
Overall structural response results from the dynamic interaction of the impact-ive lood and the structure which it impacts.
The resultant complex forcing function produces in-structure dynamic reactions in the forms of forces, moments, and shears at points away from the impactive food. As a rule, this forcing function is unknown; however, occasionally it con be estimated by incorporating knowledge of the chorocteristics of the dropped food
. (weight, size, shape, deformability), chorocteristics of the impacted structure (material properties, structural configuration), and the impoet conditions (velo-city, orientation).
The following discussion oddresses the use of energy balonce methods for the evoluotion of reinforced concrete and structural steel structures. These techniques do not require explicit knowledge of the forcing function.
The food drop methodology incorporates the conservation of energy and momentum to calculate the transmitted kinetic energy and maximum displacement to investigate the important modes of overall reinforced concrete structural- behavior. The objective of this methodology is to chorocterize structural behavior in terms of the available strain energy up to prescribed performance limits. These limits are dictated by either ductile or brittle modes of failure. The ductile mode is chorocterized by large inelastic. deflections without complete collapse, while the brittle mode may result in partial failure or total collapse. The available internal strain energy that con be obsorbed by the concrete floor system without reaching those limits of unocceptable behavior is balanced against the externally applied energy resulting from o heavy load drop.
l
, ,lt has been assumed that momentum is conserved, and the kinetic ene,rgy of the
~
drop drives the mass of the floor and induces strain. As on odditional 7
1 1
l conservatism, no credit has been token for potential sources of energy dissipo-tion through local deformation in the forms of concrete crushing and penetro-tion.
The following sections discuss specific details applicable to the evoluotion of reinforced concrete and steel structures, respectively.
Reinforced Concrete Structures Generally, the ultimate load of a concrete slob or beam system is reached prior to exceeding the hinge rotational capacity of porticular sections provided that on unstable mechanism has not formed. The hinge rotational capacity was used as a criteria to set the maximum allowable level of deflection for the concr or beam system. The hinge rotational capacity for concrete structures was developed in References 7 and 8 based on test results given in References 9 and 10 and is given os:
(7) r o = 0.0065 (d/c) d 0.07 where r u= rotational capacity of plastic hinge (radians) d = distance from the compression face to the tensile reinforcement c = distance from the compression face to the neutral oxis at ultimate strength The maximum deflection for o concrete slab or beam with a plastic hinge of its center is then given by:
(8)
Xm = (ru l/4)
- where, Xm = maximum deflection
(
,. L = span of beam 8
Rotations of the mognitude governed by equation 7 result in crocking which is Generally confined to o region below (above) the tensile reinforcement.
speaking, the section will remain intact with no crushing, spalling, or scobbing due to flexure; however, scobbing may occur os o result of shock wave motion associated with the refle. tion of tensile waves from the rear surface or shear plug formation. It has been conservatively assumed that scabbing does occur.
The load / deflection history up to the point of the ultimate loading, coupled with the maximum allowable deflection, defines the maximum level of strain energy The shear stress at obsorption provided that a shear failure has not occurred.
limiting sections was checked and compared to allowables os specified in Chapter 11 of ACI 381-77 (Reference Il).
Structural Steel Structures The maximum response of structural steel elements is determined using the commonly applied energy balance method (References I,12, and 13) by equating the externally applied kinetic energy to the available internal strain energy. The maximum permissible deflection of each structural element is given in terms of on allowable ductility ratio which is defined os:
(9)
Um Y
- Uy where Um = maximum permissible deflection Uy = deflection at the effective limit The allowable structural steel ductility ratios for impact loads have been taken from Reference I and are as follows:
t t
f e.
9
ALLOWABLE MODE OF RESPONSE DUCTILITY RATIO
- l. Flexure
- open sections 12.5
- closed sections 20
- 2. Shear ,
5
~
- 3. Compression 14 x 104 4 10 Fy ($)2 r
- 4. Torsion 0.5 tyE where Fy = minimum yield stress of the steel K = theoretical effective length factor for compression member L = length of compression member su = ultimate strain Gy = yield strain The effective yield limit corresponds to the inflection point of on equivalent closto-plastic resistance displacement curve os defined by Newmark (Reference 14). For simplicity, on equivalent elasto-plastic resistance displacement curve was developed by setting the maximum resistance equal to the octual minimum yield resistance. This procedure is conservative because it neglects the strain energy associated with the strain hardening mechanism.
Discussion of Structural Margins in addition to the conservatisms previously mentioned, the following conservo-tisms are also inherent in the methodology used in the evoluotion:
- 1. Static material strengths for concrete and steel are used.
[
Test dato shows that this property increases with the increased strain rates ossociated with dynamic loodings.
l i For example, References 13 ond 15 recommend dynamic increase factors of 1.25 for the compressive strength of concrete and 1.20 for the flexural, tensile, and compres-sive strength of structural steel.
10 l
- 2. Design (minimum) material properties for steel are used.
The onroge strength for structural steel is nearly a factor of 1.25 (Reference 11) higher than the minimum yield requirement specified by ASTM. While these foetors above minimum code strength exist and contribute to structural morgins, they are not used in the evoluotion.
- 3. Equotion 7 for hinge rotational capacity is used. This corresponds to rotations of the order of 2 degrees with minimum crocking and no crushing or scabbing. To meet necessary performance requirements (i.e., halting propa-goting failures), larger rotations in the range of 5 to 12 degrees could be tolerated. Such rotations would lead to crushing, spalling, and scabbing of the section (Reference 15); however, overall lood carrying capability is expected to remain intact. Experimental observations (Reference
- 18) suggest even further capability for well-designed and well-onchored slabs. Foilure modes at such levels initially appear to be controlled by yielding in shear and flexure followed by membrone stretching until failure occurs, normally at the support edge of the slob. Use of these larger rotational capabilities would have resulted in greater energy obsorbing copobilities of the floor system.
- 4. The analysis uses ACI 318-77 allowable shear stresses. A significant body of dato suggests the existence of higher shear copobilities. (References 18 through 26).
- 5. The structural loods are distributed directly under the dropped heavy lood. In reality, o more favorable food distribution would exist due to the load distribution copo-bility of the slab.
- 6. No credit is taken for local energy dissipation associated with any crushing of the load itself or the immediate surface of the floor.
L l
l l
l 11 e y w- g- .- - --
RESPONSES TO REQUESTS FOR IWORMATION IN SECTIONS 2.2 APO 2.3 OF ENCLOSURE 3 TO fitC DECEMBER 22,1980 Lt:Iit9t 2.2 SPECIFIC REQUIREMENTS FOR OVERWAD HAPOLING SYSTEMS OPERATING IN REACTOR BUILDING NUREG-06!2, Section 5.I.4, provides guidelines concerning the design and operation of food-handling systems in the vicinity of spent fuel in the reactor vessel or in storage. Information provided in response to this section should demonstrate that adequate measures have been token to ensure that, in this crea, 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 NUREG-0612, Section 5.I, Criteria i through 111.
ITEM 2.2-1 Identify by name, type, capacity, and equipment designator, any cranes physically capable (i.e., ignoring interlocks, moveable mechanical stops, or operating procedures) of carrying loads over spent fuel in the storage pool or in the reactor vessel.
RESPONSE: The only heavy load handling system operating in the Reactor Building that is capable of carrying loads over spent fuel in the storage pool or in the reactor vessel is the Reactor Building Crone. The Reactor Building Crane is on overhead bridge crone with a main hoist of 100 tons and on auxiliary hoist of 5 tons.
I s
12
ITEM 2.2-2 Justify the exclusion of any cranes in this area from the above .
verifying that they are incopoble of carrying heavy foods or are
- p. permanently prevented from movement of heavy loads over stored fuel or into any location where, following any failure, such load may drop into the reactor vessel or spent fuel storage pool.
RESPONSE: ~ Justification for exclusion of handling systems operating in the vicinity of the spent fuel pool is provided in the response to draf t Franklin TER Section 2.l.l.c in Enclosure I to this submittal.
a i
13
~
& v ' = , - - - , . , . ..,
ITEM 2.2-3 Identify any cranes listed in 2.2.-l, above, which you have evoluoted as having sufficient design - features to make the likelihood of a load drop extremely small for all foods to be corried and the basis for this evoluotion (i.e., complete compli-once with NUREG-0612, Section 5.1.6, or partial comptionce supplemented by suitable otternative or additional design features). For each crone so evoluoted, provide the food-handling-system (i .e., crone-lood-combination) information specified in Attachment 1.
RESPONSE: The Reactor Building crone was evoluoted to industry standards CMAA 70-1975 and ANSI B30.2-1976. It was found to meet these standards as indicated in the response to draft Franklin TER ltem 2.l.8.c in Enclosure i to this submittol.
The only cose where load handling reliability was considered was with respect to the main hoist load block cnd hook. NUREG-0612 requires that the food block and hook be considered as a heavy load. The food block is used for handling numerous loads, including the reactor vessel head, drywell head, shield plugs, and the dryer and separator units. In moving these foods, the hook, food block, rope, drum, sheave assembly, motor shafts, gears, and other load bearing members are subjected to significant stresses approaching the load rating of the crane. By
' comparison, these components are subjected to o considerably smaller load when only the hook and load block are being moved. Based on this,it is not considered feasible to postulate a rondom mechanical failure of the crane load bearing components when moving the crone load block alone.
The only feasible failure modes for dropping of the main hook and load block would be:
- 1) A control system or operator error resulting in hoisting of the block to o "two blocking" position with continued hoisting by the motor and subsequent parting of the rope (this situation con be prevented by operator action prior to "two blocking" or by on upper limit switch to terminate hoisting prior to "two blocking"); and l
- 2) Uncontrolled lowering of the food block due to failure of the holding broke to function (the likelihood of this con be made small by use of redundant holding brakes).
14 i
1 The PNPS Reactor Building crone is currently provided with one upper limit switches to interrupt power to the hoist motor prior to "two blocking." A second l upper limit switch will be odded. When power is removed, holding brakes are l outomatically applied. l
]
The holding brakes are solenoid rele.ased, and spring opplied on loss of power to the solenoid. Two holding brakes are provided for each of the main and auxiliary hoists. Eoch broke hos sufficient copocity to hold the rated load (each broke is greater than 150% of full motor torque). Additionally, inspection and maintenance procedures assure that the limit switches and holding brokes are functionc! and properly odjusted.
With the provisions described above, the two limit switches will reduce the likelihood for "two blocking" and the two holding brokes will reduce the likelihood of uncontrolled lowering of the food block. Based on these features,it is concluded that a drop of the lood block and hook is of sufficiently low likelihood that it does not require food drop analyses.
4 i
- t. .
15 i
l l
For cranes identified in 2.2.-l, above, not categorized occording l l lTEM 2.2-4 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 2.3 of this request. With respect to Criterio I through lli, provide o discussion of your evoluotion of crone operation in the Reactor Building and your
~ determination of compliance.
ITEM 2.2-4o ' Where reliance is ptoced on the installation and use of electri-col interlocks or mechanical stops, indicate the circumstances under - which these protective devices con be removed or bypassed and the administrative procedures invoked to ensure proper authorization of such oction. Discuss any related or proposed technical specifications concerning the bypass of such interlocks.
RESPONSE
There are no interlock systems or mechanical stops provided for the' Reactor Building Crone that prohibit movement of heavy loads over the spent fuel pool or reactor vessel during normal food handling operations. Safe lood paths implemented in plont procedures are relied on to prevent inadvertent movement of heavy loods over the spent fuel pool. Loads that must be moved over the' spent fuel pool or reactor vessel were oppropriately evoluoted as- ,
described in the following responses.
1 l
f 16
- _ -)
I ITEM 2.2-4b .Where rellonce is placed on the operation of the Stand-by Gas l
_ Treatment System, discuss present ond/or proposed technical
- l. specifications and administrative or phyiscal controls provided to ensure that these assumptions remain valid.
RESPONSE: Reliance is placed on secondary containment integrity and l , operation of the Stand-by Gas Treatment System for several postulated drops of heavy loads into the spent fuel pool. These postulated drops involve (1) drop of a spent fuel pool gate from above the pool and (2) drop and topple over of a refuel 4 slot plug as it is being moved along the eastern edge of the pool. The drops of interest would be following a refueling operation when freshly spent fuel is in the I pool. PNPS procedures will require that secondary containment be established
- and that one train of the Stand-by Gas Treatment System be operable when these load movements are mode.
ITEM 2.2-4c Where reliance is placed on other site-specific considerations (e.g., refueling sequencing), provide present or proposed tech-nical specifications, and discuss administrative or physical controls provided to ensure the validity of such considerations.
RESPONSE: As indicated in the following response to item 2.2-4d, reliance is placed on the sequence and timing of certain lood handling operations near the spent fuel pool.
In addition, as discussed in response to item 2.3 in this enclosure, certain load lifts are scheduled to occur only during cold shutdown conditions which limits the safety functions that must be occomplished following postulated load drops.
4 17
- - - + ,m.* .,---r
ITEM 2.2-4d Analyses performed to demonstrate compliance with Criteria i through 111 should conform to the ge'delines of NUREG-0612, Appendix A. Justify any exception token to these guidelines, and provide the specific information requested in Attachment 2,3, or 4, as appropriate, for each anofysis performed.
RESPONSE
There are three potential consequences of interest when considering food drops onto the open reactor vessel. They are: 1) loss of reactor vessel integrity, 2) fuel clodding domoge and the resultont radiological dose, and 3) fuel crushing and the possibility of a resulting criticality condition. Criterio I through lll in As Section 5.I of NUREG-0612 oddress each of these potential consequances.
indicated in Table 2, Criterio I through !!! were considerations for Regions I (reoctor vessel) and 3 (spent fuel pool area). The evoluotions below have been performed to oddress these issues.
Reactor Vessel The postulated load drops of interest onto or into the reactor vessel are (l) the reoctor vessel head, (2) moisture separator assembly and (3) steam dryer ossembly. A detailed finite element elastic-plastic analysis was performed to evoluote the effects of these drops onto the PNPS reactor vessel. These analyses and their results are described in General Electric Report No. NSEO 0982 entitled " Structural Analysis of Pilgrim / Boston Edison Company Vessel Head Drop, Shroud Head Assembly Drop, and Steam Dryer Assembly Drop Conditions -October 1982." This report was previously transmitted to NRC for review by letter dated February 28, 1983 from W. D. Harrington to Darrel G.
Eisenhut (ltr #03-59).
l The onolysis results indicate no domoge to fuel in the core or loss of vessel Integrity. Accordingly, Criterio I through 111 are met for drops onto the reactor
( vessel.
18
Spent Fuel Pool - Criterio I and 11 - Dose and Criticality Based on the safe lood paths defined by Boston Edison, the hecvy foods corried over or near the spent fuel pool that were considered for potential impact of spent fuel in the pool are (1) o shipping cask (cosk topple over of ter impocting the north edge of the spent fuel pool), (2) o spent fuel pool gate, (3) refueling slot plugs and (4) the " cattle chute" (refueling shield).
With regard to the potential for criticality as a result of fuel / rock crushing in the spent fuel pool, the generic evoluotion in NUREG-0612 was used as the basis for concluding that a criticality event is not possible. The FNPS spent fuel pool rock design is of the high density type. It relies on geometry and boral plates in the rocks to assure that fuel in the pool remains subcritical with substantial margins, during normal and obnormal conditions.
The NRC states in Section 2.2.4.2 of NUREG-0612 that,in the cose of BWR spent fuel rocks with boral poison cons:
" crushing the fuel would not significantly increase the keff of the fuel. For rocks with boral poisons, it seems inconceivable that any lood which might fall on the spent fut.1 pool would separate the fuel from the poison cons and subsequently push the ossemblies together to form o critical moss. Therefore it oppears that postulated lood drop events would not cause o criticality in a BWR spent fuel pool that uses boron plate type rocks."
On this basis, it is concluded that NUREG-0612 Criterion ll is met for postulated heavy lood drops into the spent fuel pool.
With regard to potential offsite doses, Figure 2.1-2 of NUREG-0612 was utilizied to evoluote dose consequences. The maximum number of fuel assemblies that could be domoged from o fuel pool gate drop was determined to be 9. The postulated lood drop scenario of interest is on end-on drop of the gate from opproximately I foot above the 117' el refueling floor. This corresponds to o drop of the gate when it is being returned to the pool following maintenance during o-refueling outoge. A flat or lengthwise side-on drop would not result in sufficient energy being transferred to inddie,al rods to cause loss of clodding integrity.
19
The maximum number of fuel assemblies that could be domoged from o refuel slot plug drop was determined to be 50. Figure 2.1-2 of NUREG-0612 indicates that for 50 assemblies domoged with credit for charcoal filters, on the order of 11 day decay times would be necessary to demonstrate acceptable consequences.
The postulated lood drop scenario of interest is a drop of slot plug to the ll7' el, floor odjacent to the east wall of the spent fuel pool and subsequent topple over into the pool. Both for the gate and the slot plugs, the drops of interest would be drops following completion of refueling operations. The reason for this is that when the gates and plugs are removed prior to refueling operation (potentially less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> offer shutdown), there is no newly offlooded spent fuel in the spent fuel pool. A conservative time of 14 days was used for spent fuel decoy time, corresponding to the minimum time to complete refueling operations and reach the point in the outage when the gates and plugs will be returned to their normal location. This 14 day decay time clearly exceeds the decoy time necessary to demonstrate occeptable consequences for the gate drop and exceeds the 11 day decay time for the plug drop. Accordingly, the NUREG-0612 dose criterio are met for these two load drop scenarios.
A cask topple into the spent fuel pool from the north end could result in impact of spent fuel in the storage rocks. Such a drop could demoge o substantial number of fuel ossemblies (potentially on the order of 200 assemblies). The potential offsite doses associated with domoge of this extent could potentially exceed the allowable offsite doses stated in Criterion I of NUREG-0612 depending on the decoy time of the fuel impacted. For this reason, odequate decoy time will se allowed for fuel assemblies in the region of the spent fuel pool near the cask loydown area prior to cask movement near the spent fuel pool.
The following " prerequisite" will be incorporated into lood handling procedures for casks.
20
l
" Prior to movement of the cask over or into the spent fuel pool, all spent fuel within a distance L (where L is the longest dimension of the cask) from the north edge of the pool shall have decoyed a minimum of 60 days."
With this restriction, Criterion I of NUREG-0612, Section S.l will be met for a postulated cosk drop.
With regard to Criterion I for potentici drops of the " cattle chute"into the spent fuel pool, specific load handling restrictions will be placed in the appropriate sections of plant procedures. These restrictions will sufficiently reduce the likelihood of the "cottle chute" octually dropping into the spent fuel pool to eliminate having to consider this drop scenario. The restriction will be to limit the carry height of the cattle chute to 6" obove the ll7' el. floor or obstructions such as curbs, electrico! junction boxes, etc., wha moving the food along the floor adjacent to the east wall of the spent fuel pool, and to specify that in no cose shall the " cattle chute" be corried over any portion of the pool. These restrictions will assure that, if the food is dropped while troversing along the east side of the pool, it will land on the refueling floor and remain upright, i.e.,
will not topple into the spent fuel pool. This is further assured by the fact that, (1) the " cattle chute" center of gravity is low because most of the mass is concentrated in the lead shielding of the bottom of the food (topple over, if dropped from low heights would not be expected), and (2) if dropped near the edge of the pool, impact of the pool curb would tend to rotate the "cottle chute" away from the pool. On this basis, Criterion I is not a consideration for o postulated "cottle chute" drop near the spent fuel pool.
Spent Fuel Pool - Criterion ill (Pool Intearity)
As indicated above, the loods postulated to occidentally fall into the spent fuel r pool are (1) o cask, (2) o refuel slot plug, and (3) o spent fuel pool gate. With regard to the cask, the potential for loss of fuel pool integrity hos previously been evoluoted. The results are presented in Section 10.3.6 of the PNPS FSAR.
It indicates there that onalysis has demonstrated that "domoge to the floor will l
l not result in a leakoge rate greater than the pool makeup copobility."
Accordingly, Criterion ill is met for a cask drop.
21
With regard to o postulated drop of a refuel slot plug, it is inconceivable that such a drop would result in impoet of the pool floor; the plug instead would impoet the spent fuel storage rocks. A drop of a spent fuel pool gate to the pool floor hos been analyzed. The results indicate that domoge to the floor will not result in leakoge greater than the pool makeup capability. Therefore, Criterion til is met for each of these two postulated drops.
l i
22 l
2.3 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN PLANT AREAS CONTAINING EQUIPMENT REQUIRED FOR REACTOR SHUTDOWN, DECAY FEAT REMOVAL, OR SPENT FUEL POOL COOLING l
NUREG-0612, Section 5.1.5, provides guidelines concerning the design and operation of lood-handling systems in the vicinity e' equipment or components 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 creas, either the j
likelihood of a drop which might prevent safe reactor shutdown or prohibit l
continued decoy heat removal is extremely small, or that demoge to such equipment from loods will be limited in order not to result in the loss of these safety-related functions. Cranes which must be evoluoted 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.
ITEM 2.3-1 Identify any cranes listed in 2.1-1, above, which you have evoluoted as having sufficient design features to make the i
likelihood of a load drop extremely small for all loads to be corried and the besis for this evoluotion (i.e., complete compli-once with NUREG 0612, Section 5.l.6, or partial compliance supplemented by suitable alternative or additional design features). For each crone so evoluoted, provide the food-handling-system (i .e., crone-lood-combination) information specified in Attachment I.
RESPONSE
A discussion regarding the Reactor Building crane is provided in the response to l Item 2.2 in this enclosure. The designs of the remaining handling systems l oddressed in this submittal, including the Turbine Building Bridge Crane, have 1
l also been evoluoted against oppropriate industry standards and have been found to comply with certain justified exceptions. These evoluotions are included in the response to draf t Franklin TER ltem 2.l.l.c and 2.l.8.c in Enclosure I to this I submittal.
23
As in the cose of the Reactor Building Crone, the Turbine Building Crone hoist hook and lood block have been eliminated as potential heavy lood drops on the basis of lood handling reliability. The evoluotion for the Reactor Building Crone in the response to item 2.2-3 is equally applicable to the Turbine Building Crone and accordingly, is not repeated here. In this regard, o second upper limit switch will be odded to the Turbine Building. Crone to assure adequate protection against a potential two blocking event.
ITEM 2.3-2 For any crones identified in 2.1-1 not designated as single-failure-proof in 2.3-1, a comprehensive hozord evoluotion should be provided which includes the following information:
- a. The presentation in a matrix format of all heavy loads and potential impoct areas where domoge might occur to safety-related equipment. Heavy loods 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 i
l other methods such that the impact area con be located on the plant general arrangement drawings. Figure I provides a typical matrix.
- b. For each interaction identified, indicate which of the load and impoet oreo combinations con be eliminated because of separation and redundoney of safety-related equip-ment, mechanical stops and/or electrical interlocks, or other site-specific considerations. Elimination on the basis of the aforementioned consideration should be supplemented by the following specific information:
(1) For load /torget combinations eliminated because of separation and redundancy of safety-related equip-ment, discuss the basis for determining that load drops will not offect continued system operation (i.e., the ability of the system to perform its safety-related function).
(2) Where mechanical stops or electrical interlocks are to be provided, present details showirg the areas where crone travel will be prohibited. Additionally, provide o discussion concerning the procedures that
) are to be used for authorizing the byossing of interlocks or removable stops, for verifying that interlocks are functional prior to crane use, and for f verifying that interlocks are restored to operability offer operations which require bypassing have been completed.
24 l
l
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ ___ _ _ . _ . . _ _]
(3) Where load /torget combinctions are eliminated on the basis of other, site-specific considerations (e.g.,
maintenance sequencing), provide present and/or
- proposed technical specifications and discuss administrative procedures or physical constraints invoked to ensure the validity of such considero-tions.
- c. For interactions not eliminated by the analysis of 2.3-2-b obove, identify any handling systems for specific loads which you have evoluoted as having sufficient design features to make the likelihood of a load drop extremely small and the basis for this evoluotion (i.e., complete
, compliance with NUREG 0612, Section 5.l.6, or partial compliance supplemented by suitable alternative or oddi-tional design features). For each so evoluoted, provide the lood-handling-system (i.e., crone-lood-combination) information specified in Attachment 1.
- d. For interactions not eliminated in 2.3-2-b or 2.3-2-c, above, demonstrate using appropriate analysis that domoge would not preclude operation of sufficient equip-ment to allow the system to perform its safety function following a load drop (NUREG 0 612, Section 5.I, Criterion IV). For each analysis so conducted, the follow-ing information should be provided:
(I) An indication of whether or not, for the specific load being investigated, the overhead crone-handling system is designed and constructed such that the hoisting system will retain its load in the event of seismic occelerations equivalent to those of a safe shutdown earthquake (SSE).
(2) The basis for any exceptions taken to the onolytical guidelines of NUREG 0612, Appendix A.
(3) The information requested in Attachment 4.
RESPONSE
The types of evoluotions performed for each defined load impact region are illustrated in Table 3. The evoluotion approaches and methodologies are described in the introductory sections to this enclosure.
The results of the evoluotions indicated that for the large majority of the load l
drop scenarios safe shutdown and/or decoy heat removal could be occomplished l
l 25
relying on plant safety systems to perform their normal design functions. There were, os well, several potential problems identified in some of the load impact regions. These potential problems are described below along with a description of Boston Edison's plans for resolving each.
Region 4 - North Equipment Hatch On the basis of preliminary structural evoluotion of a drop of a shipping or waste debris cask through the north equipment hatch from the 117' refueling deck elevation to the 23' el. floor, it was concluded that the floor could potentially fail. This failure could result in domoge to the torus, below elevation 23',
potentially resulting in loss of torus water inventory and subjecting safety related equipment to o radioactive steam environment should ADS operation or primary pressure relief become necessary. To address this situation, Boston Edison will as on interim measure, revise cask load handling procedures to require that the cask be lifted vertically in the northwest quadrant of the hatchway. The torus is located below the 23' el. floor in the southeast quadrant of the hatch opening. In addition, the load orientation will be such that the load impact will to the maximum extent possible pick up the load resistance capability of a major structural wall (torus compartment woll) that runs diogonally from the southwest to the northeast approximately below the center line of the 23' el. floor exposed to the hatch opening.
In the long term, Boston Edison will either (1) undertake more sophisticated structural analyses that could include:
(c) on investigation of membrone oction capoSility and higher hinge rotational capability, (b) consideration of specific load impact limiters integral to the cask, and i
(c) explicit consideration of the energy dissipation capability of the cask transport vehicle, OR 26
(2) Will design and fabricate energy obsorbing pods for temporary installation on the 23' el. floor du ing heavy lood movements in the equipment hatch.
Region 2 - Dryer / Separator Pool l Reactor Building Closed Cooling Water System (RBCCW) piping is located below the dryer / separator pool floor. The piping involved if breached could potentially result in loss of both trains of RBCCW which provides cooling support to the RHR System, Core Spray System, HPCI and RCIC. Structural evoluotions indicate that for drops of the dryer or separator from normal carry heights the .
floor will remain intact. However, for two other postulated load drops, structural analyses could not odequately demonstrate that the pool floor would protect the safety equipment below. These postulated load drops are drops of a dryer / separator pool slot plug from the 117' el. and drop of a cask from the l17' el.
to the pool floor.
Dryer / Separator Pool Plug Drop - Upon further evoluotion, it was concluded that based on handling procedures, plant configuration and load movement sequencing, the probability of a pool plug drop to the pool floor was sufficiently -
small to preclude postulating the drop.
The Dryer / Separator Pool plugs are stocked vertically. Slots in the pool walls j ore engaged by the plugs such that any drop of a plug as it is being raised or l lowered to and from the refueling floor would result in folling back down into its normal location, i.e., no impoet of the pool floor. Safe load paths are stringently f
adhered to when moving to and from their normal position. Further, dryer / separator pool plugs are typically removed and installed twice during a refueling outage, i.e., 4 movements total. For two of these movements the separator would be in its storage location of the west end of the dryer / separator pool and would protect the pool floor from direct impact from on occidental plug drop.
27 l
l
Accordingly, the only conceivable way a plug could drop to the pool floor would be for o drop to occur during one of the two lifts that the separator is not present in the pool and precisely during the lif t when the plug is not engaged in its slots, but is still over the edge of the pool. The likelihood of a drop occurring of this point is low because of (1) the small time the plugs will be in the undesired position, and (2) the structural integrity of the crone and lif ting equipment will be verified upon initial lif t off and hold of the plug, i.e., prior to the plug being in the undesired position.
Cask drop - The designated cask washdown crea for PNPS is in the dryer / separator pool. As a result of the potential problems associated with a cask drop to the pool floor, the cask washdown creo will be moved to the reactor head storage stond. This will eliminate the possibility of a cask drop to the pool floor.
Regions 6, I2 and 13 Regions 6,12 and 13 are defined by hatch openings at the 91' el, of the Reactor Building (see Table 1). The component of interest is RHR piping in the compartments below these hatch openings. The only heavy loads corried in these regions are the hatch covers themselves which weigh approximately 1500 lbs.
The possibility of hatch cover drop through a hatch opening will essentially be eliminated by appropriate procedural controls on hatch movement that will include limitations on lif t height and load orientation and a prohibition ogoinst movement of a hatch cover over on open hatch. No additional measures are judged to be necessary.
Region 7 - 117' el. Refueling Floor l
Systems evoluotions have concluded that safe shutdown and decay heat removal con be accomplished if domoge in Region 7 con be limited to the elevation immediately below the 117' el. (i.e. the 91' el.). Accordingly, structural analyses were undertaken to determine if the floor could adequately resist large load impocts such that overall floor collapse or perforation would not occur.
28
These structural analyses revealed that the floor system potentially has substantial lood impact resistance. However, premature failure of certain clip ongle connections at the ends of the steel beams supporting the floor could result in the floor system not realizing its full potential to resist the load impoet. The lood drop scenarios of interest are the' reactor vessel head and reactor cavity shield plugs.
As a result of these conclusions, Boston Edison plans to design and install modifications to the steel beam connections of interest that will allow the floor to resist load impacts of these large heavy loads from drop heights up to their normal carry height.
Region 17 - Turbine Building 51' el.
The elevations below the 51' el. turbine deck in Region l} ; Figure 11) contain electrical equipment associated with the operation of both trains of safe shutdown system components. This equipment is separated vertically (Train A at the 37' el, and Train B of the 23' el.). Conservative structural analyses were performed of this oreo to determine if domoge from postulated heavy load drops could be limited to the 37' el. immediately below the 51' el. slob. This involves demonstrating no perforation of the slab or overall collapse of the floor system.
Based on these structural analyses, it was determined that the Turbine Building 51' el. floer system had the potential for similar problems with the beam connections as described above for the Reactor Building ll7' el. However, because the weights of the loads considered for this area (see Table 2 of Enclosure I to this submittol) are considerably less than those in Reactor Building I situation, it was determined that the load drop parameters (e.g. carry height and load path) could be restricted with the implementation of procedural controls such that floor system failure would not occur. These procedural controls are j described in Enclosure I to this submittol in the response to draft Franklin TER Open item 2.l.l.c. More rigorous structural analyses of this floor system currently being performed may result in refinements to these procedural controls. If so, they will be incorporated into plant procedures.
29
I l
REFERENCES
- 1. Civil Engineering and Nuclear Power, Report of the ASCE Committee on impoetive and impulsive Loads, Vol. V, American Society of Civil Engineers, September 1980
- 2. Effects of Impact and Explosion, Summary Technical Report of Division 2, National Defense Research Committee, Vol. I, Washington, D.C.,1946 3 Vossalto, F. A., Missile impoet Testing of Reinforced Concrete Panels, HC-5609-D-l, Colspan Corporation, January 1975
- 4. Stephenson, A. E., " Full Scale Tornado Missile impact Tests," Electric Power Research Institute, Final Report NP-440, July 1977
- 5. Beth, R. A. and Stipe, J. G., " Penetration and Explosion Tests on Concrete Slobs," CPAB Interim Report No. 20, January 1943
- 6. Beth, R. A., " Concrete Penetration," OSRD-4856, National Defense Research Committee Report A-319, March 1945
- 7. ACI 349-76, Code Requirements for Nuclear Sofety-Related Concrete Structures, Appendix C "Special Provisions for Impulse and Impactive Effects," American Concrete institute,1976
- 8. Kennedy, R. P., "A Review of Procedures for the Analysis and Design of Concrete Structures to Resist Missile impact Effects," Journal of Nuclear Engineering and Design, Vol. 37, No. 2, May 1976
- 9. Mottock, A. H., " Rotational Capacity of Hinging Region in Reinforced Concrete Beams," Flexural Mechanics of Reinforced Concrete, ASCE 1965-50 (ACP SP-12), American Society of Civil Engineers,1965 30
- 10. Corley, W. G., " Rotational Copocity of Reinforced Concrete Beams,"
Journal of Structural Division, ASCE, Vol. 92, No. STS, Proc. Paper 4939, Oct.1976, pp.121-146.
I 1. Building Code Requirements for Reinforced Concrete, ACI 318-77, American Concrete Institute, December 1977
- 12. Structural Analysis and Design of Nuclear Plant Focilities, American Society of Civil Engineers,1980
- 13. " Design of Structures for Missile impoet," Topical Report BC-TOP-9A, Bechtel Power Corporation, September 1974
- 14. " Development of Criteria for Seismic Review of Selected Nuclear Power Plants," NUREG/CR-0098, N. M. Newmork and W. J. Hall for the U.S.
Nuclear Regulatory Commission, May 1978
- 15. Structures to Resist the Effects of Accidental Explosions, TM5-1300, Department of the Army, Washington, D.C., July 1965
- 16. Design of Structures to Resist the Effects of Atomic Weapons - Strength of Materials and Structuroi Elements, TM5-856-2, Department of the Army, Washington, D.C., August 1965
- 17. Personc! communication between Professor William J. Hall and Howard A.
Levin, October 5,1981
- 18. Wong, C. K. and Salmon, C. G., Reinforced Concrete Design, Intext Educational Publishers, New York,1973
- 19. Ferguson, P. M., Reinforced Concrete Fundamentals, J. Wiley, New York, 1973 1
1 31
- 20. Untrover, R. E. ond C. P. Siess, " Strength and Behavior in Flexure of Deep
~
Reinforced Concrete Beams Under Static and Dynamic Loading," Civil Engineering Studies Structural Research Series Report No. 230, University of Illinois, Urbono, October 1961
- 21. Austin, W. J., et of, "An investigot.on i of the Behavior of Deep Members of Reinforced Concrete and Steel," Civil Engineering Studies Structural Research Series No.187, University of Illinois, Urbano, January 1960
- 22. de Poiva, H.A.R., and C. P. Siess, " Strength and Behavior of Deep Reinforced Concrete Beams Under Static and Dynamic Loading," Civil Engineering Studies Structural Research Series Report No. 231, University of Illinois, Urbano, October 1961
- 23. de Poivo, H.A.R., and W. J. Austin, " Behavior and Design of Deep Structural Members - Port 3 - Tests of Reinforced Concrete Deep Beams,"
Civil Engineering Studies Structural Research Series No.1974, University of Illinois, Urbano, March 1960
- 24. Winemiller, J. R. and W. J. Austin, " Behavior and Design of Deep Structural Members - Port 2 - Tests of Reinforced Concrete Deep Members with Web and Compression Reinforcement," Civil Engineering Studies Structural Research Series Report No.193, University of Illinois, Urbano, August 1960
- 25. Newmork, N. M. and J. D. Holtiwanger, " Air Force Design Monval -
Principles and Practices for Design of Hordened Structures, AFSWC-TDR-62-138, December 1962
- 26. Crawford, R. E., et of, "The Air Force Manual for Design and Analysis of Hordened Structures," AFWL-TR-74-102, October 1974 l
1 32 1
! TABLEI l LOAD IMPACT REGIONS l
DEFIED FOR EVALUATION PURPOSES l Handling System j Region Description l 1 Reactor Vessel ,
Reactor Bldg. Crane 2 Dryer / Separator Storage Pool including Reactor Bldg. Crone elevation 74'3" below pool floor 3 Spent Fuel Pool including elevation SI' below Reactor Bldg. Crane pool floor 4 North equipment hatch at elevation 117' to Reactor Bldg. Crone hatch occess crea at elevation 23' including elevation (-)l7'6" below hatch occess creo S Mid equipment hatch at elevation 117' to Reactor Bldg. Crane hatch occess area at elevation 91'3" includ-ing elevation 74'3" below hatch occess area 6 South equipment hatch at elevation 117' to Reactor Bldg. Crane hatch occess area at elevation 91'3" includ-ing elevation 74'3" below hatch occess crea and elevation 51' below SFP demineralizer space 7 North half of refueling deck at elevation ll7' Reactor Bldg. Crone including north half of elevation 91'3" 8 South half of refueling deck at elevation 117' Reactor Bldg. Crone including south half of elevation 91'3" 9 RHR loop A monorail including area at elevo- RHR A Pump and Motor tion 23' and RHR A equipment space, elevation Monorail / Hoist
(-)l7'6" 10 RHR loop B monorail including creo at elevo- RHR B Pump and Motor tion 23' and RHR B equipment space, elevation Monorail /Holst
(-)l7'6" ll RECIRC Pump Motor Monorail including oreo Recirc Pump Motor inside drywell below elevation 37' Monorail /Holst 12 Monorail at Reactor Building elevation ll3'-3h", Spent Fuel Pool above SFP filter spoce, including area at ele- Filter Equipment votion 91'3", the SFP filter spoce at el. 74'3", Hotch Monorail /Holst and elevation 51' below SFP filter space.
-e, - . - -
.-r - - g--
l l
l l TABLEI l LOAD IMPACT REGIONS DEFIED FOR EVALUATION PURPOSES (Continued)
Region Description Handling System 13 Monorail at Reactor Building elevation ll3'-3Y2", Reactor Water Cleanup above cleon-up filter demineralizer spaces, in- Filter Equipment Hatch cluding area of elevation 91'3", elevation 74'3", Monorail /Holst below lood path and elevation 51" below clean-up filter deminerlizer spaces.
14 Reactor Auxiliary Boy Monorail, including oreo Reactor Auxiliary Boy and hatch at elevation 23', orea and hatch at Equipment Hatch elevation 3' and area at elevation (-)l7'6" Monorail / Hoist 15 Recire poinp MG Set space at elevation SI' in- Recirculation Pump ciuding corresponding area at elevation 23'. MG Set Monorail / Hoist 16 Majority of turbine deck at elevation 51' in- Turbine Building Bridge ciuding southeast equipment hatch and corres- Crone ponding areas at elevations 37',23', and -l' (also 6').
17 Load pickup area on turbine deck at elevation Turbine Building Bridge 51', northeast of turbine, including corres- Crane ponding creo at elevation 37'.
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, FIGURE 9 LOAD IMPACT REGION 8 - SOUTH REACTOR BUILDING REACTOR BUILDING - Iil' EL
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REACTOR MAtUAL SCRAM PROff CilON SY5 f[M I
REACTOR b--===========a Set >IDOWN CONTROL ROD
, DRIVE SYSTEM OFF5ITE US NO WER Ato Malt It COPOEN5ER VAILABLE i6 I
- PRF55URE ,
RELif f INil AL COOLING $ygTgg
& OLPRE55LAtl2 ATION I I CONTROLLED COOLDOWN HIGH PRE 55URE AUTOMAflC OEPRES$URIZATION "
REACTOR CORE SYSTEM E JIP t ISOLATION COOLING COOLANT NJECTION DFCI) (ADS) 0108C)
CottXNSAIE I MAN STE AM COPOEN5ER lg SUPPORI SYSTEMS
_ _ I MODE CORE RtB-LOW PRES $URE E SPRAY COOLANT HKCTION l
AUTOMATIC l l DEPRE55URIZATION
SYSTEM a l rte SHUTDOWN COOLING M(0E EXTEPOED COOtHG t
FIC'JRE 12 SAFE SHUTDOWN CONCEPT FOR 4
HEAVY LOAD EVALUATIOP6 4
FUNCTIONS SCRAM MONITORING DEPRESSURIZATION/ MAKEUP EXTEf0ED COOLNG PRESSURE RELIEF / LOW PRESSURE SHdTDOWN REACTOR AUTOMATIC COOLING MODE PROTECTION COOLANT INECTION SYSTEM DEPRES RI ATION MODE & Rm WH INSTRUMENTATION e REACTOR LEVEL e REACTOR PRESSURE 4
FIGURE 13 SAFE SHUTDOWN SYSTEMS SELECTED FOR LOAD DROP EVALUATIONS
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -- _ _ - - _ _ _ . _ - - _ _ - - - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _