ML20076F943
| ML20076F943 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/10/1983 |
| From: | Crouse R TOLEDO EDISON CO. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR 52221, 952, TAC-52221, NUDOCS 8306140500 | |
| Download: ML20076F943 (142) | |
Text
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TOLEDO
%me EDISON Docket No. 50-346
"*[";ce ua License No. NPF-3
"
- 25S5'2' Serial 952 June 10, 1983 Director of Nuclear Reactor Regulation Attention:
Mr. John F. Stolz Operating Reactor Branch No. 4 Division of Operating Reactors United States Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Stolz:
This is in response to Mr. D. G. Eisenhut's letter dated December 22, 1980 (Log No. 648) concerning " Control of Heavy Loads at Nuclear Power Plants", NUREG-0612. Attached is Toledo Edison's response to Section 2.2, 2.3 and 2.4 of Enclosure 3 (Phase II) of Mr. Eisenhut's letter for the Davis-Besse Nuclear Power Station Unit No. 1.
Very truly yourc, RPC: GAB:lah attachment (5 copies) cc: DB-1 NRC Resident Inspector
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l 8306140500 030610 PDR ADOCK 05000346 P
PDR THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO. OHIO 43652
RESPONSES TO REQUESTS FOR INFORMATION IN SECTIONS 2.2,2.3, AlO 2.4 OF ENCLOSURE 3 TO P41C DECEMBER 22,1980 LETTER 2.2 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN THE VICINITY OF FUEL STORAGE POOL NUREG-0612, Section 5.l.2, provides guidelines concerning the design and opero-tion of food-handing systems in the vicinity of stored, spent fuel. Information provided in response to this section should demonstrate that adequate measures have been taken to ensure that in this oreo, either the likelihood of a load drop which might domoge spent fuel is extremely small or that the estimated consequences of such a drop will not exceed the limits set by the evoluotion criteria of NUREG-0612, Section 5.1, Criterio I through 111.
ITEM 2.2.1 Identify by name, type, capacity, and equipment desig-notor, any cranes physically capable (i.e., ignoring interlocks, move-able mechenical stops, or operating procedures) of carrying foods which could, if dropped, land or fall into the spent fuel pool.
RESPONSE: The spent fuel cask crone (Crone H-4) is physically capable of carrying foods over spent fuel in the spent fuel pool; however, consistent with plant Technical Specification 3.9.7, various physical and administrative controls exist to prevent loads greater than 2,430 pounds (weight of one fuel assembly and its handling tool) from being carried over spent fuel in the pool. The overhead bridge and trolley crane is rated at 140 tons for the main hoist and 20 tons for the auxiliary hoist.
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g ITEM 2.2.2 Justify the exclusion of any cranes in this area from the above category by verifying that they are incapable of carrying heavy loads or are permanently prevented from movement of the hook centerline closer than 15 feet to the pool boundary, or by providing a suitable analysis demonstrating that for any failure mode, no heavy food con fall into the fuel-storage pool.
RESPONSE: The spent fuel cask crane is electrically interlocked to prevent the crone from traveling over the spent fuel pool while any load is hanging on the main hook. This interlock con only be bypassed with a key that must be obtained from the Shift Supervisor. Even upon bypassing this interlock, the main book stays inoperative; only the auxiliary hook con be used. Before any load is corried over the spent fuel pool when fuel and water are in the pool, the weights of the food must be verified to be less than 2,430 pounds and entered in the Shif t Supervisor's log. The auxiliary hook is used to move the spent fuel pool divider gates (2) which separate the spent fuel pool from the fuel transfer tube pit and the cask pit. These gates weigh 8,000 pounds; however, they are not corried over spent fuel.
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ITEM 2.2.3 Identify any cranes listed in 2.2.1, above, which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely small for all loads to be carried and the basis for this evaluation (i.e., complete compliance with NUREG-0612, Section 5.1.6 or partial compliance supplemented by suitable alternative or additional design features). For each crone so evalu-oted, provide the lood-handling-system (i.e., crane-lood-combination) information specified in Attachment I.
RESPONSE: It has not been necessary to evaluate the spent fuel cask crane identified above against the criteria of NUREG-0612, Section 5.l.6, because the handling systems themselves are not single-failure-proof.
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ITEM 2.2.4 For cranes identified in 2.2.1, above, not categorized according to 2.2-3, demonstrate that the criterio of NUREG-0612, Section 5.1, are satisfied.
Compliance with Criterion IV will be demonstrated in response to Section 2.4 of this request. With respect to Criteria i through lil, provide o discussion of your evaluation of crone operation in the spent fuel area and your determination of compliance. This response should include the following information for each crane:
Which of ternatives (e.g., 2, 3, or 4) from those identified a.
in NUREG-0612, Section 5.l.2, have been selected.
b.
If Alternative 2 or 3 is selected, discuss the crane motion limitation imposed by electrical interlocks or mechanico!
stops and indicate the circumstances, if any, under which these protective devices may be bypassed or removed.
Discuss any administrative procedures invoked to ensure proper authorization of bypass or removal, and provide any related or proposed technical specification (opero-tional and surveillance) provided to ensure the operability of such electrical interlocks or mechanical stops.
c.
Where reliance is placed on crane operational limitations with respect to the time of the storage of certain quantities of spent fuel at specific post-irradiation decoy times, provide present and/or proposed technical specifi-cations ad discuss administrative or physical controls provided to ensure thot these assumptions remain valid.
d.
Where reliance is placed on the physical location of specific fuel modules at certain post-irradiation decay times, provide present and/or proposed technical specifi-cotions and discuss administrative or physical controls provided to ensure that these assumptions remain valid.
e.
Analyses performed to demonstrate compliance with Cri-teria i through 111 should conform to the guidelines of NUREG-0612, Appendix A. Justify any exception taken to these guidelines, and provide the specific information requested in Attachment 2, 3, or 4, as appropriate, for each analysis performed.
RESPONSE: As indicated in the response to item 2.2.2, physical, administra-tive, and Technical Specification controls exist to prevent carrying heavy loads over spent fuel in the spent fuel pool. Nonetheless, a spent fuel pool divider gate could potentially impact the fuel in the unlikely event that it is dropped and falls in on unfavorable orientation.
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For this reason, alternative 4 of those identified in NUREG-0612, Section 5.l.2, hos been selected to demonstrate compliance with Criteria i through Ill.
Structural, radiological release, and criticality eviouations were conducted consistent with the guidelines of NUREG-0612, Appendix A, and Attachments 2 and 3 to the NRC's December 22,1980 request for information.
l The steel spent fuel pool divider gates weigh approximately 8,000 pounds and are 25 feet long,2 feet wide, and 4 inches thick. The gates are stored just odjacent to their normal positions within the two slots dividing the spent fuel pool from the fuel transfer tube pit and the cask pit. At the maximum lif t height, the j
gates are carried approximately 42 feet above the pool floor and 28 feet above the top of the spent fuel, i
The gate was assumed dropped in various configurations, each resulting in a different area of exposure for the spent fuel. As a first iteration, the number of fuel rods potentially damoged was determined, and a conservative estimate was made of the resulting offsite doses assuming that all impacted fuel was damaged.
The basic assumptions for the dose calculations are contained in the Davis-Besse Final Safety Analysis Report (FSAR) Section 15.4.7, " Fuel-Handling Accident" and, specifically in Section 15.4.7.2, " Accident Analysis-Accident Outside Con-tainment"(Reference 41). This section states explicitly that: "The assumptions and guidelines of Safety Guide 25 were used in this analysis" and, further, provides the conservative and plant-specific assumptions in FSAR Table 15.4.7-l. These assumptions were utilized in the analysis rather than those in NUREG-0612. We believe the analysis to be extremely conservative because of fuel assemblies in the pool were considered to have characteristics identical to the " hottest assembly" of Table 15.4.7-l. Based on on average power density of 15.661 MW (FSAR Table 4-10), the operational power level of 27.9 MW for this t
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" hottest assembly" equates to o simply calculated peaking factor of opproxi-mately 1.78.
Doses were computed by simply taking a ratio of the number of domoged fuel rods under this scenario over the number assumed in the FSAR and_ multiplying times the doses documented in FSAR Table 15.4.7-2. In this manner, and not i
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considering the energy of the drop or the capability of water drog and the assemblies and rocks themselves to reduce this energy, :t was determined in this preliminary and very conservative calculation that the guidelines of NUREG-0612 would be exceeded for o postulated drop with the largest surface area projection; however, the dose values remained well within 10 CFR 100 regulatory limits (Reference 3).
Finally, drop energy calculations were performed, realistically considering the drog effect of the water and the energy obsorbing capability of each fuel rod. It was determined that the limiting dose would be to the whole body and that IS assemblies represented the maximum number that could be domoged without exceeding NUREG-0612 guideline doses. Calculations were then performed to determine if various drop scenarios could potentially lead to domoge of greater than IS assemblies, taking into occount water drog and the energy-obsorbing copobility of the fuel. Structural evoluotions demonstrated that a value of 19.8 ft.-pounds per fuel rod is on oppropriate energy obsorbing limit below which the fuel cladding integrity could be assumed to be intact (see response to item 2.3).
For the most limiting drop geometry, less than IS assemblies were predicted to be domoged.
Based upon this evoluotion, we conclude the dose guidelines of NUREG-0612 will not be exceeded for postulated drops of the spent fuel pool gates onto spent fuel.
In the criticality evoluotion, the assumptions of NUREG-0612, Section 2.2, were found to be compatible with the Davis-Besse design. The Davis-Besse spent fuel pool is described in FSAR Section 9.l.2, " Spent Fuel Storage." This section states that the center-to-center fuel assembly spacing of 21 inches in either direction is "... sufficient to maintain a Keff of 0.90 or less." Based on the fact that the original design analyses conservatively considered spacing alone, with no credit for borated water, on evoluotion was performed, in accordance with NUREG-0612 Section 2.2, to determine the Keff under worst-case fuel domoge conditions, taking credit for o boron concentration of 1,800 ppm.
(System l
Procedure SP l104.42, Spent Fuel Pool Operating Procedure, requires that the i
boron concentration in the spent fuel pool be maintained equal to or greater than 1,800 ppm). The evoluotion included the assumption that all fuel to be impacted had an enrichment of 3.5 weight percent, which is the " highest probable onrichment" according to FSAR Table 4-5.
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3 Using Figure 2.2-3 of NUREG-0612, it is evident that the increases in reactivity due to fuel crushing (represented in terms of a decreasing water /U02 volume ratio) are insignificant, since the Davis-Besse water /UO2 rotto of 1.22 (1/0.82 from FSAR Table 4-3) is near the peak of the 2,000 ppm neutron multiplication factor curve. Therefore, decreasing the water /UO2 ratio leads to less reactivity.
Thus, considering the negative reactivity of the boron in the pool, crushing of the fuel will not cause Keff to increase above 0.90 which satisfies the NUREG-0612 guidelines.
A structural evaluation was completed to ossess the potential for perforation, scabbing, and leokoge of the 5' thick spent fuel pool (SFP) base slab due to a postulated 42' drop of a SFP divider gate. The energy dissipated by drog in the l
SFP water was conservatively neglected. The worst configuration for potentio!
perforation of the pool f!cor was assumed neglecting the structural resistance of the steel liner plate. The methodology and criterio documented in the response to item 2.3 were utilized. It was concluded that perforation and scabbing of the pool floor were not probable. Furthermore, penetration of the floor slab was predicted to be insignificant. Based upon this evaluation, leakage of the pool resulting from o postulated SFP divider gate drop from its maximum carry height is not expected.
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2.3 SPECIFIC REQUIREMENTS OF OVERHEAD HANDLING SYSTEMS OPER-ATING IN THE CONTAINMENT NUREG-0612, Section 5.l.3, provides guidelines concerning the design and opero-tion of lood-handling systems in the vicinity of the reactor core. Information provided in response to this section should be sufficient to demonstrate that adequate measures have been taken to ensure that in this area, either the likelihood of a load drop which might damage irradiated fuel is extremely small or that the estimated consequences of such a drop will not exceed the limits set by the evaluation criterio of NUREG-0612, Section 5.1, Criterio I through lli.
l ITEM 2.3.1 Identify by name, type, capacity, and equipment designa-for any cranes physically capable (i.e., taking no credit for any interlocks or operating procedures) of carrying heavy loads over the reactor vessel.
RESPONSE: The containment polar crane (Crone H-5) and the reactor service crane (Crone H-ll) are physically capable of carrying heavy loads over the reactor vessel. The containment polar crane is rated at 180 tons for the main hoist and 25 tons for the auxiliary hoist. The reactor service crane is a traveling type crane with rolls located on the D-ring (steam generator, pressurizer, and I
l reactor coolant pump enclosures) walls and is rated at 5 tons; however, it is not j
presently being used. There are plans to modify this crane for future use at
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which time heavy load handling operations will be evaluated.
Plant procedures restrict load lif ts over the reactor vessel when loaded with fuel except for the reactor vessel head, the reactor plenum assembly, the reoctor missile shields, reactor testing equipment, and the associated lif ting devices.
While the handling system has been evaluated and found to comply with the intent of applicable industry standards, possessing demonstrated reliability and margins to failure, the consequences of a postulated handling accident involving a drop of any of the above loads have been evoluoted. The results of these evoluotions are documented in the response to item 2.3.4.
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ITEM 2.3.2 Justify the exclusion of any crones in this area from the above category by verifying that they are incapable of carrying heavy foods, or are permanently prevented from the movement of any load either directly over the reactor vessel or to such a location where in the event of any food-handling system failure, the food may land in or on the reactor vessel.
RESPONSE: The reactor service crane is not presently being used. Heavy load handling operations will be evoluoted in the future when this crone is put into service.
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i ITEM 2.3.3 Identify any cranes listed in 2.3.1 above which you have evoluoted as having sufficient design features to make the likelihood of a load drop extremely small for o!! loods to be corried and the basis for this evoluotion (i.e., complete compliance with NUREG-0612, Section 5.I.6, or partial compliance supplemented by suitable alternative or additional design features). For each crone so evolo-oted, provide the load-handling-system (i.e., crane-lood-combination)
Information specified in Attachment I.
RESPONSE: As indicated in the response to item 2.2.3 above, it has not been found to be necessary to evaluate the Polar Crane against the criteria of NUREG-0612, Section 5.l.6.
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ITEM 2.3.4 For cranes identified in 2.3.1 above not categorized according to 2.3.3, demonstrate that the evoluotion criteria of NUREG-0612, Section 5.1, are satisfied. Compliance with Criterion IV will be demonstrated in your response to Section 2.4 of this request. With respect to Criteria I through lil, provide a discussion of your evaluation of crone operation in the containment and your determination of compliance.
This response should include the following information for each crane:
ITEM 2.3.4.0, Where reliance is placed on the installation and use of electrical interlocks or mechanical stops, indicate the circumstances under which these protective devices can be removed or bypassed and the administrative procedures invoked to ensure proper authorization of such action. Discuss any related or proposed technical specifica-tions concerning the bypassing of such interlocks.
RESPONSE: The polar crane load block has not been included in any of the heavy load drop evaluations described in subsequent responses for the reasons given below:
NUREG-0612 requires that the load block and hook be considered as a hecyy load.
The load block is used for handling numerous loads. In moving these loads, the hook, load block, rope, drum, sheave cssembly, motor shafts, gears, and other load bearing members are subjected to significant stresses approaching the food rating of the crane.
By comparison, these components are subjected to a considerably smaller food when only the hook and load block are being moved.
Based on this, it is not considered feasible to postulate o random mechanical failure of the crane load bearing components when moving either the main hoist or auxiliary hoist load block without a load.
The only two 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 porting of the rope (this situation con be prevented by operator action prior to "two blocking" or by on upper limit switch to terminate j
hoisting prior to "two blocking"); and 2)
Uncontrolled lowering of the load block due to failure of j
the holding broke to function (the likelihood of this con be made small by use of redundant holding brakes).
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The Davis-Besse polar crane main and auxiliary hoists or provided with upper and lower limit switches which, when the hoist block reaches a predetermined limit of travel, will interrupt the current to the hoist motor. Also, the polar crone is equipped with a Revere digital weight indicator and limiter, providing the capability to limit the maximum load lifted and, therefore, imported to the lifting devices. This food cell gives o digital readout to the crane operator at the controller, and also terminates hoisting outomatically and alarms if the upper limit set-point is reached. The use of such a device is beyond the requirements of CMAA or ANSI, and significantly reduces the likelihood of domoge to the crane or lifting devices due to on overload.
The main and auxiliary hoists are each equipped with a dual electric broke system. The brakes are solenoid released and spring applied on loss of power to the solenoid. Each broke is designed to sustain full load independent of the other broke. The broke is rated at greater than 150 percent of the motor full load torque.
With the provisions described above, the limit switches will reduce the likelihood for "two blocking" and the redundant holding brakes will reduce the likelihood of uncontrolled lowering of the load 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 load drop onalyses.
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ITEM 2.3.4.b.
Where reliance is placed on other, site-specific consi-i derations (e.g., refueling sequencing), provide present or proposed technical specifications and discuss administrative or physical controls provided to ensure the continued validitiy of such considero-tions.
RESPONSE: Plant procedures and Technical Specification 3.9.5 require that direct communication be maintained between the control room and personnel inside contginment at the refueling station.
Additionally, plant procedures require that direct communication be maintained between the crane operator and the ground crew. These provisions provide overall assurance that heavy load handling operations are safely followed and that mitigative action could be promptly initiated in the unlikely event that it is required.
Plant procedures restrict load lif ts over the reactor vessel when loaded with fuel, except for the reactor vessel head, the react'or plenum assembly, the reactor missile shields, reactor testing equipment, and the associated lifting devices. Furthermore, plant procedures provide for the following:
1.
The reactor vessel head carry height over the reactor vessel flange is limited to the height necessary to clear the guide studs (3'-3 3/8").
This is accomplished by raising the head to this elevation and then moving it north and away from the reactor vessel within the refueling canal prior to being lif ted to elevation of the head storage stand.
2.
The potential drop height of the reactor plenum assembly over the core is physically limited and corresponds to the indexing fixture height (6'-lh"). This physical limitation exists because perfect alignment of the plenum assembly within the indexing fixture is required to slide the plenum in and out of the reactor vessel.
3.
The missile shields (6) are removed and replaced in a sequence that precludes a direct impact onto the reactor vessel in the unlikely event that a drop should occur. This is accomplished by lifting each of the interior (those closest to the reactor centerline) missile shields until they just clear their holddown studs, and then translating them north (or south) over adjacent missile shields prior to east (or west) translation for laydown on top of the D-rings.
These procedures ensure that for all potential drop sce-narios over the reactor that a dropped shield would eitheT fall directly back on the D-ring support ledge or onto on adjacent shield.
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Each of the above provisions is governed by written procedures and is strictly enforced by operators in charge of lif ts using the polar crane.
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ITEM 2.3.4.c Analyses performed to demonstrate compliance with Criterio I through ill should conform to the guidelines of NUREG-0612, Appendix A.
Justify any exception taken to these guidelines, and provide the specific information requested in Attachment 2, 3, or 4, as appropriate, for each analysis performed.
RESPONSE: There are three potential consequences of interest when consider-ing postulated load drops onto the open reactor vessel. They are: 1) loss of reactor vessel integrity and obility to either cover the core or deliver water to the core through vessel penetrations, 2) fuel cladding damage and the resultant radiological dose, and 3) fuel crushing and the possibility of a resulting criticolity condition. Criteria 1 through lil in Section 5.1 of NUREG-0612 oddress each of these potential consequences. The following evoluotions have been performed to address these issues.
The general steps used to perform the evolvations are outlined below:
1.
Identification of heavy loads including a full chorocteriza-tion of the load weight, dimensions, materMI 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 behov-ior. For reactor internals mechanical elements and fuel, identify on a component-by-component basis the potential failure mechanisms (i.e., bending, shear, buckling, etc.).
4.
Incorporating I through 3 above. nrovide early input to the systems evoluotions to factor structural information into systems evoluotions assumptions.
5.
Conduct detailed structural evoluotions following bound-ing systems, dose, and criticality eviavations when these initial approaches foil to demonstrate acceptable conse-quences. The detailed structural evoluotions include:
a.
Specification of impact energy considering, as appropriate, the energy dissipated due to the tron-sfer of momentum, fluid drog, bouyoney, "dosh pot" effect of the internois sliding within the reactor, etc.;
b.
Model development for assessing dynamic response utilizing empirical dato os necessary; 15 i
c.
Development of failure criteria based upon stability, leak tightness, or fuel cladding strain considero-tions; d.
Computation of the strain energy absorbed prior to reaching the prescribed performance limits; e.
Assessment of structural response end structural consequences of drop.
The structural evoluotion methodology and criteria follow the guidelines of NUPEG-0612, Appendix A, and recommendations made by the American Society of Civil Engineers Technical Committee on Impulse and Impact Loads (Reference 7).
Details of the structural evoluotion are provided in Section 3.1 of Appendix B.
On the basis of this evoluotion, it is concluded thct Criteria i through lli are met for all postulated drops over the reactor vessel.
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t 2.4 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN PLANT AREAS CONTAININC EQUIPMENT REQUIRED FOR REACTOR SHUTDOWN, CORE DECAY HEAT REMOVAL, OR SPENT FUEL POOL COOLING t
'NUREG-0612, Section 5.l.5, provides guidelines concerning the design and opera-tion of load-handling systems in the vicinity of equipment or components l
rsquired for safe reactor shutdown and decay heat removal.
Information provided in response to this section should be sufficient to ' demonstrate that adequate measures have been taken to ensure that,in these areas, either the likelihood of a load drop which might prevent safe reactor shutdown or prohibit continued decay heat removal is extremely smali or that damage to such equipment from ioad drops will be limited in order not to result in the loss of these safety-related functions. Cranes which must be evaluated in this section have been previously identified in your response to 2.1-1 and their loads in your response to 2.l.3.3.
ITEM 2.4.1. identify any cranes listed in 2.1.1 above, which you have evaluated as having sufficient design features to make the likeflhood of a 16ad drop extremely small for all loads to be carried and the basis for this evaluation (i.e., complete compliance with NUREG-0612, Section 5.l.6, or partial compliance supplemented by suitable alterantive or additional design features). For each crane so evalu-ated, provide the load-handling-system (i.e., crane-load-combinction) information specified in Attachment 1.
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RESPONSE: The handling systems of interest listed in response to item 2.1.1 are the Containment Polar Crane, the Spent Fuel Cask Crane, the Component-Cooling Water Pump Monorails, and the Service Water intake Structure Gantry Cranes. !t has not been necessary to evaluate any of these handling systems against the criteria of NUREG-0612, Section 5.1.6, because the handling systems themselves are not single-failure-proof.
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ITEM 2.4.2 For any cranes identified in 2.1-1 not designated as single-failure-proof in 2.4-1, o comprehensive hozord evoluotion should be provided which includes the following information:
1.
The presentation in a matrix format of all heavy loads and potential impact creas where damage might occur to safety-related equipment.
Heavy loads identification should include designation and weight or cross-reference to information provided in 2.I-3-c. Impact areas should be identified by construction zones and elevations or by some other method such that the impact area con be located on the plant general arrangement drawings. Figure i pro-vides a typical matrix.
2.
For each interaction identified, indicate which of the load and impact area combinations con be eliminated because of separation and redundancy of safety-related equip-ment, mechanical stops and/or electrical interlocks, or other site-specific considerations.
Elimination on the basis of the aforementioned considerations should be supplemented by the following specific information:
a.
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).
b.
Where mechanical stops or electrical interlocks are to be provided, present details showing the areas where crane travel will be prohibited. Additionally, provide o discussion concerning the procedures that are to be used for authorizing the bypassing of interlocks or removable stops, for verifying that interlocks are functional prior to crone use, and for verifying that interlocks are restored to operability of ter operations which require bypassing have been completed.
c.
Where food /torget combinations are eliminated on the basis of other, site-specific considerations (e.g.,
maintenance sequencing), provide present ond/or
- proposed technical specifications and discuss admin-istrative procedures or physical constraints invoked to ensure the continued validity of such considero-tions.
3.
For interactions not eliminated by the analysis 2.4-2-b, obove, identify any handling systems for specific foods which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely small and the basis for this evoluotion (i.e., complete 18 1
compliance with NUREG-06!2, Section 5.l.6, or partial compliance supplemented by suitable alternative or addi-tional design features).
For each crane so evaluated, provide the food-handling-system (i.e.,
crone-lood-combination)Information specified in Attachment 1.
4.
For interactions not eliminated in 2.4-2-b or 2.4-2-c, obove, demonstrate using appropriate analysis that damage would not preclude operation of sufficient equip-ment to cllow the system to perform its safety function following a load drop (NUREG-0612, Section 5.I, Criterion IV). For each analysis so conducted, the following infor-motion should be provided, o.
An indication of whether or not, for the specific load being investigated, the overhead crane-handling system is designed and constructed such that the hoisting system will retain its load in the event of seismic occelerations equivalent to those of a safe shutdown earthquake (SSE).
b.
The basis for any exceptions taken to the analytical guidelines of NUREG-0612, Appendix A.
c.
The information requested in Attachment 4.
RESPONSE: For the four handling systems listed in 2.4.1 above, a combination of systems and structural evoluotions was utilized to determine if Criterio 111 and IV of NUREG-0612 are met for all postulated load drop scenarios. An overview of the systems and structural evoluotions for creas inside and outside contain-ment follows. Details of the systems and structural evaluations are provided in Appendices A and B, respectively.
The heavy foods are listed in Tables I through 4 for loads carried by the containment polar crane, the component cooling water pump monorails, the spent fuel cask crane, and the intoke gantry crane, respectively. Also shown on Tables I through 4 are the approximate weights, applicable operating procedures, and lif ting equipment associated with each heavy load.
Each heavy load was evoluoted to determine the locations where it is handled on both a routine and nonroutine basis (e.g., major repairs). Table 5 presents the results of this evoluotion for heavy loads handled by the polar crane for each of j
nine regions inside containment. These nine regions are identified'in Figures l 19
through 9, Appendix A, by the shaded portion of the diagram. These regions are based upon the configuration of the containment building and load handling procedures. For Region I, Table 5 notes whether the load is handled with the reactor vessel head on or removed.
For areas inside containment, plant procedures provide the following restric-tions:
1.
The polar crane is not operated over the refueling canal when any fuel assembly is being moved.
2.
During the period when the reactor coolant system is pressurized above 300 psig and is chove 200 F, and fuel is in the core, the polar crone hoists are not operated over the D-rings.
3.
Movement of loads is limited to the minimum height required to clear the area and a direct path to the laydown area.
This minimizes the time and height required for each lift.
All of the above provisions are governed by written procedures and are strictly enforced by operators in charge of lifts using the polar crane.
The handling locations for each handling system outside containment are as follows:
o Component cooling water pump monorails - creas direct-ly under each monorail (see Figure 10, Appendix A);
o Spent fuel cask crane - area within boundary of spent fuel pool. (Note: the cosk crane is interlocked such that the crone cannot travel over the spent fuel pool when the crane is in the cask handling mode).
o Intake gantry crone -- crea of travel for crane hook in vicinity of the service water pumps and piping (see Figure 11, Appendix A).
The specific approach chosen to evaluate load drop consequences is based upon the completeness of available information, a preliminary assessment of the likelihood of success of possible opproaches, and the estimated level of resources 20 T'
required to perform the necessary evaluations.
To help select a preferred evaluation approach, Table S was utilized to identify controlling heavy loads and consequences of interest for each region.
The weight, size, and handling procedures are important considerations in this selection. Table 6 provides a summary of the identified controlling heavy loads and consequences of interest.
The objective of the " systems evaluations" was to demonstrate that safe shutdown and long-term cooling could be achieved, assuming that certain combinations of equipment were lost due to damage resulting from o postulated load drop.
For drop scenarios evaluated inside containment, it was assumed that:
1.
If the reactor head was installed, the reactor was depres-surized to about 300 psig on decay heat removal, and cooled to less than 200 F (see Procedure SP l104.46, Polar Crane System Procedure (Reference 4);
2.
If the reactor head was removed, the reactor was in " Cold Shutdown" or " Refueling" as defined in the Technical Specifications.
For drop scenarios outside containment, it was assumed that the reactor could be in any operational mode, including full power operation.
The steps used to perform the systems evoluotions inside containment are outlined below:
1.
Identify the safety functions required to achieve or main-toin safe shutdown.
2.
Identify the plant systems required to accomplish those safety functions and potential backup systems.
3.
Select a region for consideration.
4.
Identify the equipment within the region that could poten-tiolly be lost if a load drop were to occur in that region.
5.
Assume the equipment within the region is losi, and determine the resultant effects of the loss of this equip-ment on the' safety functions required to achieve safe 21
shutdown.
In order to identify which combinations of systems / equipment failures result in a loss of core cooling capability, o set of event trees was developed as a tool.
6.
If the evaluation indicates safe shutdown and/or core cooling capabilities still exist even assuming loss of the equipment in the selected region, then no further analyses are required for the region.
7.
If the evaluation indicates a failure to demonstrate safe shutdown or adequate core cooling with the systems in the region assumed lost, then structural analyses are per-formed to attempt to show that damage ccn be localized, or alternative approaches (e.g., procedural encnges, modi-fications) must be considered.
The steps used to perform the systems evaluations outside containment near the component cooling water pumps and the service water intake structure are outlined below.
l.
Identify the system (s) of interest in the area.
2.
Identify the components (of the systems of interest) potentially damaged in the area. Figures 10 and 11 of Appendix A show important equipment in the vicinity of the component cooling water pumps and service water pumps, respectively.
3.
Determine if the system functions would not be lost due to the loss of the component identified.
4.
If the evaluation reveals that the system function would not be lost, no further analysis is required.
If the evoluotion reveals that the system function could be lost, then consider alternatives such as load path or lifting restrictions, or determining if damage can be limited or restricted to a smaller area.
The general steps used to perform the structural evaluations are outlined below:
l.
Identification of heavy loads, handling systems, and hand-ling locations including a full characterization of the food weight, dimensions, material properties, and structural i
characteristics.
2.
Development of postulated drop scenarios based upoo i
reclistic consideration of plant procedures.
22
3.
Review of important structural engineering aspects of impacted structural elements to fully characterize behavior.
For reinforced concrete and steel elements, identify drops which control " local" response (e.g., pene-tration, scabbing, spalling, perforation, etc.); loads that control "overall" structural response (e.g., large inelastic deformations or obrupt failures of principal structural members, etc.); and/or loads that may induce behavior that exhibits combined response such that either overall or local fai!vre modes would control.
4.
Incorporating I through 3 above, provide early input to the ' systems evaluations to factor structural information into systems evoluotions assumptions.
5.
Conduct detailed structural evaluations following bound-ing systems, dose, and criticality evaluations when these initial opproaches fail to demonstrate acceptable conse-quences. The detailed structural evaluations include:
a.
Specification of impact energy considering, as op-propriate, the energy dissipated due to the transfer of momentum, fluid drag, bouyancy, " dash pot" effect of the internals sliding with the reactor, etc.;
b.
Model development for assessing dynamic response utilizing empirical data os necessary; 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 consequences of drop.
The following discussions present results, conclusions, and recommendations for Regions I through 9 inside containment and for the component cooling water pump crea, the service water intoke structure area, and the spent fuel pool area, respectively.
Region I - Reactor Vessel This region was addressed in the response to item 2.3.4.c.
23
Region 2 - North End of Operating Deck - 603' Elevation The bounding postulated load drop for Region 2 is a drop of the reactor vessel head as it is being placed on its storage stond. Structural evoluotions predicted excessive deformations of the elevation 603' floor resulting from this drop.
Accordingly, the systems evoluotions assumed that equipment in the entire area below Region 2 was lost. It was shown that the decay heat removal system would be available and that the emergency core cooling system was not required.
Based upon these evoluotions, it was concluded that core cooling would be maintained.
Region 3 - D-Ring Enclosures The controlling postulated load drops in Region 3 are the missile shields and reactor coolant pump motors. Structural evoluotions demonstrated that the consequences of a missile shield drop on top of either D-ring are acceptable.
Similar evoluotions were not undertaken to assess the consequences of a reactor coolant pump motor drop due to the complexity and cost of the required evoluotions. Accordingly, the systems evoluotions assumed that as a conse-quence of the motor drop, the decay heat removal system was lost for cases including a reactor coolant system break and no reactor coolant system break.
The loss of various instrumentation and piping important to safety was also assumed. (See Appendix A, Section 2.1.3 for a detailed discussion). It was shown that requisite portions of the emergency core cooling system would be available or that, if required, o systems alignment including the PORV would be available.
Based upon these evoluotions, it was concluded that core cooling would be maintained.
Region 4 - Refueling Canal (North End)
The bounding postulated load drop for Region 4 is a 76' drop of a missile shield onto the refueling cavity floor.
A drop of the plenum assembly was also considered.
Structural evoluotions showed the consequences of the plenum assembly drop to be acceptable; however, the consequences of a missile shield drop were found to be unocceptable due to excessive deformations. Accordingly, l
I i
24 n
e-
i the systems evaluations assumed a loss-of-coolant occident in one loop and i
damage to one train of core flood piping, in neither case was core cooling capability lost. The effects of leakage or flooding resulting from perforation of the refueling canal were also evoluoted and found to be acceptoble with respect j
to equipment survivability.
Region 5 - Refueling Canal (South End)
The bounding postulated load drop for Region 5 is a plenum assembly drop on the cavity floor. Structural and systems evaluations found the consequences of this drop to be insignificant and, therefore, acceptable.
Region 6 - Piping Enclosures Region.6 is potentially impacted by secondary impacts resulting from damage due to perforation or scobbing of the elevation 653' slab above.
Structural j
evaluations demonstrated significant protection of this region; notwithstanding, the systems evaluations assumed a loss of one line of decay heat removal injection (east piping enclosure) and the decay heat removal alternative pressuri-zer spray line. The decay heat removal function was shown to be available in the redundant train.
Based upon these evoluotions, it was concluded that core cooling would be maintained.
Region 7 - Grating in Southeast Quadrant - 603' Elevation Region 7 is potentially impacted by secondary impacts resulting from damage j
due to perforation or scabbing of the elevation 653' slab above. Structural evaluations showed that a drop of an in-core instrument tank access hatch cover could cause such scabbing. Accordingly, the systems evolutions conservatively assumed the loss of the decay heat removal suction piping, various instrumer:f a-tion, and power cables. (For o detailed discussion, see Appendix A, Section 2.1.9).
It was shown that requisite portions of the emergency core cooling system would be available. Based upon these evoluotions, it was concluded that core cooling would be maintained.
25
~
Region 8 - In-Core instrument Area The bounding postulated drop is a drop of the in-core instrument tank access hatch cover from its position at the hatch penetration at elevation 653' onto the elevation 606' concrete slab which is over the in-core instrument tank. Struc-tural evaluations showed that such a drop would not perforate the elevation 606' slob; however, it was conservatively assumed that a hatch cover could fall directly through a rectangular opening in the in-core instrument tank through o E-inch checkered steel plate cover. The systems evaluations conservatively assumed that all instrument tubes inside the tank would be severed, cousing a loss-of-coolant occident. It was shown that requisite portions of the emergency core cooling system would be available.
It was concluded that core cooling would be maintained.
Region 9 - Area Adjacent to the Equipment Hatch The bounding postulated load drop is a drop of a reactor coolant pump motor under extreme circumstances when a motor is taken through the main equipment hatch for major repairs. A specific structural evaluation' was not completed; however, the systems evoluotions assumed a loss of the make-up and purification system piping cousing a loss-of-coolant accident.
It was shown that the emergency core cooling system would be unaffected. Based upon these evoluo-tions, it was concluded that core cooling would be maintained.
Component Cooling Water Pump Area The bounding postulated load drops for loads handled by the component cooling water pump monorails are the pumps themselves. The consequence of concern is that a pump could fall in such a manner os to directly domoge operable portions of the component cooling water (CCW) and service water (SW) systems or that CCW ond SW piping beneath the floor could be domoged by scobbing. Structural evaluations have shown that scabbing is not probable; furthermore, the effects of scobbing would be insignificant.
The systems evoluotions have shown that adequate physical separation exists between electrical and mechanical compo-nents such that redundant parts of each system would not be jeopardized. Based 26
upon these evaluations, it was concluded that the CCW and SW functions would remain operable for postulated handling accidents during operation.
Service Water intake Structure Area l
The bounding postulated food drops for loads handled by the intake gantry crane are the service water pump motors. The consequence of concern is that a motor could fall directly down its hatch as it is being hoisted or fall over the valve room area as it is being transported away and damage operable portions of the service water system. Structural evoluotions have shown that perforation and scabbing of the volve room roof is not probable; however, the systems evalua-tions have assumed the possibility of scabbing. These evaluations have shown that adequate physical separation exists between electrical and mechanical components such that redundant parts of the service water system would not be jeopardized. Based upon these evaluations, it was concluded that the service water system function would remain operable for postulated handling accidents during operation.
Spent Fuel Pool Area The bounding postulated load drops for loods handled by the spent fuel pool cask crane are the spent fuel pool divider gates. The consequence of concern is that a drop could damage the floor of the pool such that it leaks and causes a loss of pool inventory. Structural evaluations have shown that the pool integrity will not be compromised by a postulated drop.
Based upon the above evoluotions, it is concluded that Criteria Ill and IV of NUREG-0612 are met for all postulated food drop scenarios.
27
TABLEI POLAR CRAE NAVY LOADSl APPROX.
APPLICABLE WEIGHT OPERATING LIFTING LOAD (POUNDS)
PROCEDURES EQUIPMENT l.
Reactor Plenum Assembly 2 l19,000 SPl 104.465 Plenum Assembly SPl505.01 7 Lifting Rig 2 2.
Reactor Vessel Head 3 330,000 SPI 104,46 Reactor Vessel SPl504.016 Head Lif ting Rig 3.
Internals indexing Fixture 4 31,100 SPI 104.46 Plenum Assembly Lifting Rig 2 4.
Plenum Assembly Lifting Rig 18,500 SPI 104.46 N/A SPl505.01 5.
Automatic Reactor 32,000 SPI iO4.46 Slings and Shockles inspection System (ARIS) 6.
l-Beam D-Ring Groting 12,000 SPI 104.46 Wire Rope Slings Supports and Shockles 7.
Steam Generator Removable 14,100 spi l04.46 Wire Rope Slings Supports Supports and Shockles 8.
Reactor Missile Shields (6) 94,500 eo.
M-73 Reactor Missile SPI 104.46 Shield Lifting Horness 9.
Polar Crane Load Block II,200 SPI 104.46 N/A
- 10. Steel Working Plotform 2,700 SPI 104.46 Wire Rope Slings and Shockles I1. 20" Steam Generator Snubbers 7,000 SP I IC'4.46 Wire Rope Slings and Shockles
- 12. Irradiated Specimen Cask 6,000 SPI 104.46 Wire Rope Slings l
and Shockles
-r
TABLEI (continued)
APPROX.
APPLICABLE WEIGHT OPERATING LIFTING LOAD (POUNDS)
PROCEDURES EQUIPMENT
- 13. Letdown Coolers 5,000 SPI 104.46 Wire Rope Slings and Shackles 14 Equipment Hatch Covers (Region 2)
SPI 104.46 Wire Rope Slings o 603' el. - 4 covers for I hatch 10,000 ea.
o 585' el. - 2 covers for o hatch 32,000 ea.
- 15. Cere Flooding Tank Hatch 8,000 eo.
SPI 104.46 Wire Rope Slings Covers for each of 2 hatches and Shockles (Regions 2 and 9)
- 16. In-Core Instrument Tonk 6,800 ea.
SPl 104.46 Wire Rope Slings Access Hatch Covers at and Shockles 603' el. (Region 8)
- 17. Motor Removal Hatches at 5,000 ea.
SPI 104.46 Wire Rope Slings 603' el. - I cover for each and Shackles of 2 hatches (Region 2)
- 18. Plenum Assembly Stand 6,000 SPI 104.46 Wire Rope Slings and Shockles
- 19. Core Support Barrel Stand 6,500 SPI 104.46 Wire Rope Slings and Shackles
- 20. Reactor Vessel Head 12,000 SPl 104.46 N/A Lif ting Rig SPl504.01
- 21. Reactor Coolant Pump (RCP) 4,000 SPI 104.46 Wire Rope Slings Rotating Element and Shackles
- 23. Reactor Coolont Pump 84,000 SPI 104.46 Wire Rope Slings and Shackles
- 24. Core Support Assembly 224,000 SPI 104.46 Core Support SPl50S.01 Assembly 2S. Reactor Cavity Seal Ring 3,000 SPI 104.46 Wire Rope Slings and Shackles
TABLEI (continued)
APPROX.
APPLICABLE WEIGHT OPERATING LIFTING LOAD (POUNDS)
PROCEDURES EQUIPMENT
- 26. Refueling Canal Walkways (2) 2,700 SPI 104.46 Wire Rope Slings and Shackles 1.
For reference, a heavy load is defined as the weight of a fuel assembly plus its handling tool; 2,430 lbs.
2.
The total weight of the Plenum Assembly lift includes:
the Plenum
- Assembly, and the Plenum Assembly Lif ting Rig. This lifting rig includes the following components: Head and Internals Handling Fixture; Head and Internals Handling Extension; and the Internals Handling Adapter, Pendants, and Spreader Ring.
3.
The total weight of the Reactor Vessel head lift includes: the head; the service structure; studs, nuts, and washers; and the Reactor Vessel Head Lifting Rig. This lifting rig includes the following components: Head and Internals Handling Fixture, Head and Internals Handling Extension, two turnbuckle pendants, and three head lifting cables.
4.
The total weight of the Internals Indexing Fixture includes: the indexing Fixture and the Plenum Assembly Lifting Rig.
5.
SP l104.46, " Polar Crone System Procedure."
6.
SP IS04.01, " Reactor Vessel Closure Head Removal and Replacement."
7.
SP 1505.01, " Reactor Internals Removal and Replacement."
TABLE 2 COMPOENT COOLING WATER PUMP MONORAll (3) IEAVY LOADS APPROX.
APPLICABLE WEIGHT OPERATING LIFTING LOAD (POUNDS)
PROCEDURES EQUIPMENT Component Cooling Water 5,400 SPI 104.131 Slings and Pumps Shackles Component Cooling Water 4,800 SPI 104.13 1 Slings and Pump Motors Shackles 1.
SPI 104.13, " Component Cooling Water Pumps Monorail System Procedure,"
includes sufficient discussion to olert maintenance personnel to the safety concern. In addition, it contains sufficient prerequisites and precautions to assure that the operator is qualified, that the monorail, lifting equipment, and hoist have been inspected and maintained, that the lifting equipment selected is of sufficient capacity, and that proper system isolation has been effected.
1 TABLE 3 SPENT FUEL CASK CNI KAVY LOADS APPROX.
APPLICABLE WElGHT OPERATING LIFTING LOAD (POUNDS)
PROCEDURES EQUIPMENT Pool Divider Gates 8,000 SPI 104.502 Wire Rope M-S43 Slings and Shockles 1.
This crane is used to handle loads in the Equipment and Fuel Handling Area and for foods around the spent fuel pool. It will eventually be used to handle spent fuel casks. Loads lifted in the Equipment and Fuel Handling Area are not corried near the spent fuel pool and, therefore, are not listed in this table.
2.
SP 1104.50," Spent Fuel Cask Crane Operating Procedure.a 3.
Maintenance Instruction M-54, Removol/ Replacement of Gate Between the Spent Fuel, New Fuel Storage / Transfer Pools.
l
i TABLE 4 lNTAKE GANTRY CRAE KAVY LOADSI APPROX.
APPLICABLE WEIGHT OPERATING LIFTING LOAD (POUNDS)
PROCEDURES EQUIPMENT l.
Service Water Pump 7,800 SPI 104.531 All intake Gontry Crone heavy loads are lif ted with wire rope slings
~
ond shackles 2.
Service Water Pump Motor 8,600 SPI 104.53 3.
Cire Water Makeup Pump 5,700 SP1104.53 4.
Makeup Pump Motor 3,500 SPI 104.53 5.
Roof Top Hatch Covers 2,800 SP i l04.53 6.
Dilution Pump 9,500 SP l 104.53 7.
Dilution Pump Motor 3,000 SP1104.53 8.
Diesel Fire Pump 5,200 SPl 104.53 9.
Diesel Fire Pump Motor 3,500 SPI 104.53
- 10. Screen Wash Pump 3,560 SPI 104.53 II. Screen Wash Pump Motor 1,300 SPl l04.53 1.
SPI 104.53, "Intoke Gantry Crane System Procedure."
l TABLE S POLAR CRAE FEAVY LOADS HAPOLED BY REGION INSIDE CONTAINMENT APPROXIMATE REGION (FROM FIGURES I THROUGH 9)
WEIGHT LOAD (POUNDS)l J*
2 3
4 S
f 7
8 9_
l.
Reactor Plenum Assembly 2 Il9,000 o#
c c
a a
e e
e c
2.
Reactor Vessel Head 3 330,000 o
a e
a e
e e
c c
3.
Internals indexing Fixture 4 31,100 o#.
c b
o e
e c
c c
4.
Plenum Assembly Lifting Rig 18,500 a#
c a
a a
e c
e c
S.
Automatic Reactor inspection System 32,000 a#
c c
o a
e c
c a
(ARIS) 6.
l-Beam D-Ring Grating Supports 12,000 e
c b
c c
c c
c c
i 7.
Steam Generator Removable Supports 14,100 e
c b
c c
c c
c c
8.
Reactor Missile Shields (6) 94,500/ea.
a+
c a
a c
c e
c c
9.
Polor Crane Load Block 11,200 a+# a a
a a
e c
o a
- 10. Steel Working Platform 2,700 e
e a
o a
c c
c c
f I1. 20" Steam Generator Snubbers 7,000 c
b b
c c
c c
c b
- 12. Irradiated Specimen Cask 6,000 e
a e
o a
e c
e a
- 13. Letdown Coolers S,000 c
b c
c c
c c
c b
I
TABLE 5 (continued)
I
)
APPROXIMATE REGION (FROM FIGURES I THROUGH 9)
WElGHT LOAD (POUNDS) l*
2 3
4 5
6 7
8 9
I4. Equipment Hatch Covers (Region 2) c b
c c
c c
c c
c e at 603' el. - 4 covers for I hatch 10,000/eo.
e at 585' el. - 2 covers for I hatch 32,000/eo.
- 15. Core Flooding Tonk Hatch Covers -
8,000/eo.
c b
c c
c c
c c
b 4 covers for each of 2 hatches (Regions 2 and 9)
- 16. in-Core instrumentation Tank Access 6,800/eo.
c c
a c
c c
a o
e Hotch Covers at 653' el. (Region 8)
- 17. Motor Removal Hatches at 603' el.
5,000/eo.
c b
c c
c c
c c
c I cover for each of 2 hatches (Region 2)
- 18. Plenum Assembly Stand 6,000 e
c c
b o
e c
c c
- 19. Core Support Barrel Stand 6,500 c
e a
b b
c c
c c
- 20. Reactor Vessel Head Lifting Rig 12,000 a+
o a
o e
e c
e c
- 21. Reactor Coolant Pump (RCP) 4,000 c
b b
c c
c c
c b
Rotating Element
- 22. RCP Motor 102,000 e
b b
b b
c c
c b
- 23. Reactor Coolont Pump 84,000 e
b b
b b
c c
c b
l
- 24. Core Support Assembly 224,000 o#
c c
b b
c c
c c
- 25. Reactor Covity Seal Ring 3,000 o#
c c
o e
c c
c c
- 26. Refueling Canal Walkways (2) 2,700 o+
c o
o c
c c
c c
TABLE 5 (continued)
KEY a - lood normally handled in region b - load potentially handled in region under nonroutine circumstances (e.g., major repairs) c - lood not expected to be handled in region FOOTNOTES
+ - RPV head on For Region I:
l
- - RPV head removed l
I. For reference, the weight of a fuel assembly plus handling tool is 2,430 lbs.
- 2. The total weight of the Plenum Assembly lif t includes: the Plenum Assembly and the Plenum Assembly Lifting Rig. This lifting rig includes the following components: Head and Internals Handling Fixture; Head and Internals Handling Extension; and the Internals Handling Adopter, pendants, and Spreoder Ring.
- 3. The total weight of the Reactor Vessel Head lift includes: the head; the service structure; studs, nuts and washers; and the Reactor Vessel Head Lifting Rig. This lif ting rig includes the following components: Head and internois Handling Fixture, Head and Internals Handling Extension, two turnbuckle pendants, cnd three head lif ting cables.
- 4. The total weight of the Internals Indexing Fixture includes: the Indexing Fixture and the Plenum Assembly Lif ting Rig.
6
TABLE 6
SUMMARY
OF CONTROLLING WAVY LOAD DROPS Ato POTENTIAL CONSEQWNCES OF INTEREST BY REGION INSIDE CONTAINMENT Ato LOCATIONS OUTSIDE CONTAINMENT POTENTIAL CONTROLLING CONSEQUENCES REGION HEAVY LOAD OF INTEREST I
Missile Shields Radiological release due to (Reactor Head On) fuel failures resulting from loads imported through con-trol rod drive mechanisms l
Reactor Vessel Head Damage to reactor coolant (Reoctor Head Off) loops and core flood piping resulting in a loss of core cooling capability Plenum Assembly Radiological release due to fuel failures 2
Reactor Vessel Head /
Damage to core flood /LPl/
Reactor Coolant Pump PORV piping 3E,3W Reactor Coolant Pump Damage to reactor coolant pressure boundary, domoge to DHR suction, damage to RCS and SG instrumentation 4
Missile Shields Damage to reactor coolant pressure boundary, damage to core flood piping Plenum Assembly Flocding due to loss of cavity intes rity (when lifted with cav!ry flooded)
S Plenum Assembly Flooding due to a loss of cavity integrity
~
t
TABLE 6 (continued) l POTENTIAL CONTROLLING CONSEQUENCES REGION HEAVY LOAD OF INTEREST 6E Scobbing due to o drop Domoge to auxiliary pressuri-at elevation 653' zer spray piping and LPl/DHR piping 6W Some os Region 6E None 7
Scobbing due to drop of Domoge to DHR piping in-Core instrumentation Tonk Access Hatch Covers 8
In-Core Instrumentation
~ Domoge to reactor coolant Tonk Access Hatch Covers pressure boundary, domoge to in-core instrumentation (e.g.,
core thermocouples) 9 Reactor Coolont Pump Domoge to make-up and (Remote) purification piping ARIS Tool (Routine)
Component Cooling Component Cooling Domoge to component cool-Water Pump Area Water Pump Motor ing water piping, power to pump motors Spent Fuel Pool Spent Fuel Pool Gates Domoge to fuel (dose and cri-Area ticality considerations)
Service Water Service Water Pump Domoge to service water pip-Intake Structure ing, power to pump motors Area
i APPEIOlX A l.0 SYSTEMS EVALUATION INTRODUCTION AND METHODOLOGY As part of the evoluotion of heavy load handling operations at Davis Besse, o number of postulated load drops inside containment was addressed by performing
" systems evoluotions." in addition, two other areas outside containment, the area around the component cooling water pumps and the service water intoke structure, were addressed by systems evoluotions.
The objective of these systems evoluotions was to assess the capability of the plant to safely shut down and maintain a safe shutdown condition as a result of load drops which were assumed to domoge equipment within certain prescribed regions. This typically involved determining if the primary core cooling mode could be lost and, if it could, determining if backup cooling modes would be available. In order to perform the evoluotions, it was necessary to identify the plant functions required to reach and maintain core cooling. The core cooling functions are dependent upon initial plant conditions and, for some areas inside containment, these functions are also dependent upon the possibility of a reactor coolont system (RCS) break due to domoge os a result of the load drop assumed.
Nine load drop creas (regions) inside containment were identified based on the configuration of the containment building and specific locations where heavy loads are typically corried. The nine regions are identified in Figures I through 9 by the shaded portion of the diagram. A number of scenarios were developed for load drops inside containment based upon normal operating procedures. These scenarios (coses) were then analyzed with event trees that displayed the systems / equipment required to accomplish the core cooling functions.
It was then necessary to identify the equipment that could potentially be lost if a load drop were to occur in the regions inside containment. This involved a number of assumptions regarding the possibility of damaging equipment; these assumptions were based on the potential loads which could be dropped in a region and, in some cases, physical protection which may be available.
A-l
As the lost step, the resultant effects of the lost or damaged equipment on the functions required to maintain adequate core cooling were determined.
1.1 Plant Conditions For the analysis of potential load drops inside containment, it was assumed that if the reactor head was installed, the reactor was depressurized to about 300 psig, on decay heat removal, and cooled to less than 200 F (see procedure SP l104.46, Polar Crane Operating Procedure (Reference 4). For the evoluotions with the head removed, the plant was either assumed to be in " Cold Shutdown" or
" Refueling" as defined in the Technical Specifications. The possibility of a reactor coolant system break was evoluoted for each region which contains RCS piping.
For potential drops in the component cooling pump area, the spent fuel pool area, and the service water intoke structure, it was assumed that the reactor could potentially be at any operating condition, including full power operation.
1.2 Safe Shutdown Systems 1.2.1 Inside Containment To establish the safe shutdown systems of interest, it was necessary to define existing plant methods which con be utilized to perform core cooling.
For all regions inside containment, it was assumed that prior to the postulated load drop, core cooling was being accomplished by the Decoy Heat Removal (DHR) System. Based on Emergency Procedure EP 1202.32," Loss of Decay Heat Removal Emergency Procedure" (Reference 6), backup methods could be utilized to maintain core cooling in the event of loss of the DHR system. These methods were included in the evoluotions, as appropriate. In addition, other possible methods not in EP 1202.32 were investigated. These methods were then also included in the evoluotion, if appropriate.
l A-2 i
For the methods investigated, see Section 1.3.3 which presents the event trees for each case onelyzed. For cases (Case I A, IB, IC) where there is a potential RCS break, it was assumed that DHR would be unusable. For these cases, the primary cooling mode utilized some method of injection of water; low pressure injection (LPl), makeup (MU), or high pressure injection (HPI) from the borated water storage tank (BWST), and then recirculation from the sump in the longer term. These methods are described in EP 1202.32.
For the case where the reactor vessel (RV) head is on and no RCS break assumed (Case 2), if DHR is lost, the backup method (described in EP 1202.32) is to repressurize the RCS and utilize natural circulation with the steam generators for cooling. This mode requires makeup capability, the pressurizer (heaters and sproy), and feedwater.
Another possible backup for these conditions would be to supply water to the RCS with the HPI, LPI, or MU pumps and to remove the water and heat through the power operated relief volve (PORV). After a length of time, the BWST would be emptied and recirculation from the sump would be needed.
The above backup methods could also be utilized in case of a small RCS break with the RV head in place.
For the cose when the RV head is removed, and no RCS break is assumed (Case 3), if DHR were lost, it would be possible to inject water with LPI from the BWST. When the BWST volume is depleted, a flow path through the fuel transfer tubes to the spent fuel pool (SFP) con be established with the DHR pumps returning the water to the vessel. This method is described in EP 1202.32.
The SFP cooling con be continued with the SFP cooling system.
l As another part of the systems review, instrumentation for indicating core cooling was evoluoted.
The instrumentation of interest inside containment l
included the RCS parameters of pressure, temperature, and level. In the cases where the steam generators were utilized, steam generator level indication was also evoluoted.
A-3
l I.2.2 Other Areas Outside Containment l
To establish the safe shutdown systems of interest, we considered only those systems in the vicinity of the handling system of concern. For the component cooling water (CCW) pump area, this involved only the component cooling water (CCW) system itself and the service water (SW) system. For the Service Water intoke structure, this involved only the SW system.
The equipment layout is shown in Figures 10 and lI for the areas under
{
consideration in the vicinity of the component cooling water pumps and the service water pumps, respectively.
l.3 Steps in the System Evoluotion 1.3.1 Assumptions Regarding Loss of Equipment The loss of equipment was evaluated in a conservative manner using the following ossumptions:
(l)
For load drops inside containrnent, the loss of DHR was investigated as a first step. In addition to this direct loss of the DHR system, it was conservatively assumed that if a LOCA could occur, the DHR would become inoperable due to loss of inventory and recirculation from the sump would be required.
(2)
All equipment in a given region (at all elevations) was initially assumed lost.
(3)
If RCS piping or connecting piping was in a region, on RCS pipe break was assumed to occur, and its effect on core cooling was evaluated ossuming the simultaneous loss of other equipment that could be impacted in the region.
If on isolation valve (normally closed volve or check valve) was located between the RCS and the drop region, o break was not assumed. In addition, potential for RCS pipe crimping was evaluated for ottoched RCS cooling lines.
(4)
In some cases, electrical cabling for motor operated volves would be predicted to be impacted. it was assumed that if a load drop disables the cabling, the MOV will fail os is. The circuit protection will operate os necessary, A-4
ond unless a large unisolable pipe break could occur concurrently and prevent access to the volve, manual operation of the volve could still be performed.
(S)
With respect to piping, it was assumed that if a pipe was in a region of concern, then it was lost, unless it could be shown to be safe from domoge based on some other consideration, such as distance from the octual load impact point.
(6)
With regard to pipe breaks (other than the RCS), the offected system or subsystem was assumed lost unless the capability to isolate, repair, or plug the break could be shown.
(7)
Loss of electrical control and interlock signals, other than for required indication, were not investigated.
It was assumed that if power feeds to large motors (such as pumps) remained intact, then manual operation at switch-gear or motor control centers could be performed if automatic control and interlock signals were. lost.
(8) instrumentation circuits were investigated to verify that if any indication of interest could be lost, redundant indication would be available.
Typical verification involved tracing two redundant circuits of interest to verify that, for the regions involved, at least one would not be impacted.
1.3.2 Completion of the Tables - Region-by-Region inside Containment Tables showing the equipment assumed lost have been completed for each region utilizing the above assumptions.
The results for Regions I through 9 are displayed in Tables I through II. The completion of each table included the following steps:
(1)
Identify the systems of interest. For the initial conditions assumed, this always included the Reactor Coolant Sys-tem (RCS) and the Decoy Heat Removal (DHR) System.
(2)
Identify the components (of the systems of interest) potentially domoged within the region under considero-tion.
(3)
Determine if the system function (e.g., core cooling) could be lost due to the loss of the domoged components.
A-S
(4)
Determine if the component may be undomoged due to o more realistic analysis (i.e., more localized damage justi-fied).
1.3.3 Event Trees for Drops inside Containment in order to identify which combinations of systems / equipment failures could potentially result in a loss of core cooling capability, a set of event trees was developed. These event trees cover five cases that could be encountered for load drops inside containment. They are:
Case I A - Reactor Vessel Head Removed - Load Drop Results in on Unisoloble RCS Pipe Break Case IB - Reactor Vessel Head in Ploce - Load Drop Results in a Small Unisolable RCS Pipe Break Case IC - Reactor Vessel Head in Place - Load Drop Results in a Large Unisolable RCS Pipe Break Case 2 - Reactor Vessel Head in Place - Load Drop Does Not Result in on Unisolable RCS Pipe Break Case 3 - Reactor Vessel Head Removed - Load Drop Does Not Result in a Unisolable Reactor Coolant System (RCS) Pipe Break The event trees for these cases are displayed in Figures 12 through 16.
The event trees, for the most part, identify success and failure paths at the system level. For any particular load drop, the success or failure of a particular system was evoluoted by determining whether any of the components required for operation of that system and located inside containment could potentially be domoged by the load drop.
If such components could be damaged, then a determination was made os to whether loss of that system component could result in loss of the system function. Once the success or failure of the system of interest for each case was determined, the path on the event trees corresponding to the particular load drop event being postulated could be identified.
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in all cases of a RCS break (I A, IB, IC), it was initially assumed that DHR would be unusable. Therefore, DHR does not appear on these event trees. in the other cases, if loss of core cooling was predicted, the possibility of loss of the DHR system was reonalyzed in a more realistic manner. If the path for a particular drop scenario corresponded to successful maintenance of core cooling (indicated by the term "OK), then no further evaluation of that drop scenario was required.
if the path culminated with an asterisk, then alternative core cooling modes were considered, i.e., cooling modes other than those included ir, the event trees.
The paths taken for each cose in each region are displayed in Tables 12 through
- 22. A discussion of those paths and alternative cooling modes as appropriate are presented in Section 2.1.
l.3.4 Steps in the Systems Approach - Areas Outside Containment For the CCW pump area and the SW Intake Structure, the following steps were performed:
(1)
Identify the system (s) of interest in the area.
(2)
Identify the components (of, the systems of interest) potentially damaged in the area.
(3)
Determine if the system function could be lost due to the loss of the component identified.
(4)
If the evaluation reveals that the system function would not be lost, no further analysis is required. If evaluation reveals that the system function could be lost, then consider alternatives such as load path or lifting restric-tions, or determining if damage con be limited or restrict-ed to o smaller area.
2.0 RESULTS AND CONCLUSIONS 2.1 Containment - Containment Polar Crane 2.1.1 Evoluotion of Region 1 - Reactor Vessel Structural evoivations of postulated missile shield and reactor vessel head drops
~
'onto the reactor vessel have shown that tbc vessel nozzle integrity would not be A-7
\\
w
lost. Dropping of the reactor vessel head has been determined, by structural analysis, to bound the missile shield drops with regard to vessel damcae. This potential problem (head drop) has been avoided by the development of procedures requiring laterol movement of the RV head, at minimum distance above the vessel flange, both prior to ascent on removal and offer descent on replacement.
Nonetheless, RCS nozzle breaks and one core flood nozzle break are assumed, as are cases involving no RCS breaks. In no case is core cooling determined to be lost.
The equipment identified in Table I (and the accompanying notes) was assumed to be lost as a result of a load drop on Region I.
Also indicated on Table I is whether or not the equipment failures are predicted to result in loss of the system function. The notes explain the system failure conclusions.
The conclusions regarding the system failures were then used to enter the event trees applicable to the postulated load drop. The opplicable event trees are those for Cases I A, IB, IC, 2, and 3. For the event tree ossessment, refer to Table 12.
For Case I A, none of the flow paths from the BWST through LPI or HPI or MU into the vessel, followed by recirculation through the sump and LPI, are predicted to fail. Therefore, successful core cooling is represented by Path I on the event tree. However, even if the BWST hos been drained to fill the refueling canal offer the head is removed, water from this canal will flow through the RCS break to the sump. This is depicted by success Path 4.
With regard to Case IB, the PORV and recirculation (sump and LPI) con continue to cool the core in the event of a small RCS break; Path I defines success.
Case IC considers a large RCS break with the RV head in place. Cooling con be accomplished via Path I, utilizing the some equipment as in Case I A.
Case 2 involves no RCS break, with the RV head in place. Becaust there is no RCS domoge and DHR is intact, Path I represents success.
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Case 3 also involves no RCS break, but with the RV head removed. Since DHR is not affected, success is represented by Path I.
2.1.2 Evaluation of Region 2 - North End of Operating Deck -603' Elevation Although there are several potential load drops in this Region, including the RV head, hatch covers, RCPs, and SG snubbers, there is a limited amount of equipment of importance to core cooling in the area where these loads are normally handled.
Core flood tank l-2 and its associated DHR/LPI injection line are located in this Region. A LOCA would be prevented by a check valve on the injection line, one of two on this line, and located close to the reactor vessel and outside this Region.
The lYa" pressurizer spray line, the 3" pressurizer PORV line to the quench tank and the pressurizer level instrumentation are all located in this Region outside the east D-ring. There are no identified loods that are handled in this oreo (elevation 653' north of the east D-ring) of the Region. However, on analysis was performed assuming the loss of this equipment. Although a LOCA could occur, success was achieved through Path I in the event tree for Case IC For Case IB, it was conservatively assumed that a small break could occur and the PORV would not be available; therefore, on alternative cooling mode had to be considered. It was determined that DHR would still be usable (small break in high point of system would not render DHR unusable) and could be used in conjunction with makeup.
Cases 2 and 3 were also analyzed. Since loss of DHR is not predicted in this Region, Path I represents success for both cases. With regard to the loss of all pressurizer level indication, o reevaluation of the location of the transmitter and circuits revealed that they would be protected by floors above.
Table 2 identifies the pertinent equipment and associated systems, and Table 13 summarizes the event tree assessment.
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i 2.l.3 Evaluation of Region 3E - East D-Ring Enclosure l
The entire "A" loop of the RCS, including the pressurizer and its two code safety valves, is contained within this enclosure, os are the single DHR suction line and the single discharge from the MU&P system (in octuality, MU flow is discharged through one of the HDI lines in this enclosure).
l Potentiai load drops in this Region include a missile shield and a reactor coo'ont pump motor (if removed for maintenance).
Conservative and yet realistic ossumptions were made regarding the consequences of load drops. For example, a reactor coolant pump motor drop was assumed that could disable, by crimping of piping, both the DHR suction line and the MU&P discharge line.
The concurrent loss of the PORV, the piping for which is located on the opposite side of the enclosure (the volve itself is in Region 2), was not assumed concurrently.
As shown on Table 3, all cases were considered for this Region. Although a system loss (MU&P) wo assumed, the maintenance of core cooling was not lost, and the overall results for this Region were satisfactory.
The loss of all pressurizer level indication would most likely occur only in a case which also would cause a RCS break at the high point in the system (i.e., the pressurizer). In such a case, the level indication would not he necessary. For the event tree assessment, refer to Table 14.
2.l.4 Evoluotion of Region 3W - West D-Ring Enclosure This region contains loop "B" of the RCS and the suction for the MU&P system.
As in Region 3E, the applicable loads are the missile shields and a reactor coolant pump motor (if removed for maintenance).
All cases were reviewed for this region similar to Region 3E, and in each a success path was determined.
Table 4 presents the opplicable systems of interest. For the event tree ossessment, refer to Table 15.
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i 2.1.5 Evoluotion of Region 4 - Refueling Canal (North End)
The loods of concern in the Region are the missile shields, the RV head, the plenum assembly and, potentially, the reactor coolant pump motors. The missile shields are the controlling loads. Also, the RV head is lif ted over the north end of the cavity. The plenum assembly may be lif ted with the cavity flooded or dry.
Analysis of the systems aspects of a load drop into this Region has shown that there is the potential for a LOCA in one loop. (The probability of impacting both loops is considered extremely remote because of the physical separation). There also is the chance of causing a LOCA by impacting one of the two core flood injection lines downstream of the second check volve, which is near the RV. In neither case is core cooling lost.
Another concern in this region involves the leakage or flooding resulting from perforating the refueling canal liner and base as a result of the load drop. An evaluation was performed, assuming the entire contents of the RCS, one core flood tank, and the " upper level" volume of the refueling canal were released I
through the RCS break and " hole" in the canal. The amount of water was less than that released during a LOCA, so equipment survivability was judged not to be a problem.
All cases were reviewed for this region. Table 5 and the accompanying notes present the results of the review. For event tree assessment, refer to Table 16.
2.l.6 Evaluation of Region 5 - Refueling Canal (South End)
The load of interest for this region is the plenum assembly. The only cause for concern in this region was flooding, because no cooling equipment of concern could be hit by the potential load drops. Table 6 presents the results of the flooding analysis regarding the environmental qualification of safety-related equipment. For event tree assessment, refer to Table 17.
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9 2.1.7 Evoluotion of Region 6E - East Piping Enclosure There is no identified food drop in this particular region; however, a conservative analysis was performed. Although the DHR/LPI injection line that ties in to core flood tank l-2 is routed through this oreo, neither the DHR letdown line nor the other DHR/LPI injection line to core flood tank l-l is offected. Also, although the DHR otternative pressurizer spray line is located inside this region, its loss does not have adverse impact because the spray system is not needed.
Tables 7 and 18 present the results of the systems evoluotion for this region.
2.1.8 Evoluotion of Region 6W - West Piping Enclosure As in Region 6E, there is no plausible drop of any load in this region; however, a conservative analysis was performed.
Tables 8 and 19 present the results of the systems evoluotion.
2.1.9 Evoluotion of Region 7 - Groting in Southeast Quadrant - 603' Elevation This region contains equipment for several systems that function during shut-down, extended cooling of the reactor core, or that may serva os backups during the contemplated scenarios. The equipment includes:
1)
DHR suction piping 2) 2 HPI pipes (one of which corries makeup flow) 3)
LPl/DHR injection to piping from core flood tank 1-2 4)
MU through one HPl pipe 5)
DHR alternative pressurizer spray from DHR pump l-2 6)
Substantial instrumentation and power cabling, including:
a) pressurizer level instrumentation b)
S.G.A level instrumentation A-12
c)
RCS temperature - cold leg wide range d)
RCS pressure e)
Core thermocouples There is the possibility of storage of the in-core instrument tank hatch covers at elevation 653' over this region.
Therefore, there would be a possibility of scabbing down from 653' if a cover were dropped. Nonetheless, on evaluation was performed assuming a load drop. Although much equipment could be " lost" with the potential drops in this region, analysis has shown that continued core cooling will not be lost because of the redundancy and location of equipment. As on example, the most difficult case to analyze was Case 2 (no RCS break; RV head in place). Backup cooling mode I would be disabled because loss of the DHR pump l-2 alternative spray line would result in inability to utilize the pressurizer for natural circulation. However, the power cable to the PORV would not be affected because of its location. Thus, backup cooling mode 2 (BWST and HPI or LPI or MU; PORV; Recire) could be utilized.
The results of the Region 7 evoluotion are presented in Table 9. For the event tree assessment, see Table 20.
2.'I.10 Evaluation of Region 8 (In-core Instrument Area)
Region 8 includes the in-core instrument tank, which houses 52 tubes that are at reactor pressure and are connected to a pressure boundary rated at 2,500 psig and located within the tank itself. Procedure EP 1202.32 notes that each tube could release approximately 42 gallons per minute if severed at low system pressure. Assuming all are severed by the potential load drop, there could be o total break flow of 2,184 gallons per minute, which is within the makeup copobility of the ECCS (3,000 gallons per minute per LPI system).
The DHR suction piping is also located in this Region and is considered " lost" for all plausible lood drops, as is the makeup function of the MU&P system.
The most important component is the emergency sump, located in close proximity to, but below, the in-core instrument tank. The controlling drop is a A-13
hatch cover from the 653' elevation down the hatch and into the in-core instrument tank. This could cause damage to the tank and associated equipment.
The potential of a hozord to the sump was then investigated. The location of the tank and sump, the protective screens and grotes for the sump, and the structural evaluation of the load drop all mitigate against there being any significant damage to the sump or the two DHR/LPI suction lines.
4 Case 3 considered such a lood drop with the RV head removed and no RCS break.
This analysis considered on improbable drop covering the entire Region and thus destroying both DHR suction and the alternative path utilizing the fuel transfer tubes. In reality, there is no such drop that could cause this. Therefore, the
" alternative cooling mode" for Case 3 is the DHR system, which would remain intact if a drop were to damage the fuel transfer tubes.
With regard to instrumentation, it was predicted that all core thermocouple information could be lost for a drop into the instrument tank. In addition, a LOCA could occur. Core cooling could still be ochieved as shown in the event trees. Pressurizer level would still indicate core coverage.
Table 10 presents the results for this region. For the event tree assessment, refer to Table 21.
l I
2.l.11 Evaluation of Region 9 - Area Adiocent to Equipment Hatch - SW j
Quadrant - 603' Elevation The controlling load drop for this region would be the ARIS tool or, in unusual circumstances, an RCP motor. Although containing a substantial amount of safety-related equipment, this region represents no challenge to the maintenance of core cooling because: (l) DHR would not be offected in this region; and (2) although impacting MU&P system piping can result in a LOCA, the emer-gency sump would not be affected.
Table 11 and ottoched notes present the results for this region. For the event tree assessment, refer to Table 22.
1 A-14
2.2 Component Cooling Water Pump Room Monorails The component cooling water (CCW) pump room was reviewed to determine if a CCW pump or motor, when being lif ted and removed on its monorail, could fall in such a manner to damage operable portions of the CCW system, the service water (SW) system, or the CCW and SW piping beneath the floor of this room.
The CCW system is comprised of two (100%) redundant cooling loops. The three CCW pumps are interconnected with isolation volves to allow the third pump to replace either Pump I or 2.
The three pumps and cross connect volving are located in the CCW pump room. Service water piping which serves the CCW heat exchangers is located below the pump room floor. Also, the pump motor cabling is routed beneath the pumps.
The potential effects of dropping a pump motor were analyzed based on the equipment layout (see Figure 10). Pump l-2 is at the west end of the CCW room. The center line of Pump l-2 is ll' from the center line of Pump l-3, which in turn is ll' from the center line of Pump l-l at the east end of the room.
It was determined that a drop of either CCW end pump (i.e., Pump l-l or 1-2) would damage piping associated with its own cooling loop and could, at the most, damage one set of redundant isolation valves (either CC-4 and S for Pump 1 or CC-6 and 7 for Pump 2), and that the other independent loop would remain operable because of the isolation provided by the other valves. However, with only one CCW loop operable, Technical Specification 3.7.3.1 requires that the second loop be made operable within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If it cannot be made operable, the plant must be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
The drop of pump CCW Pump l-3 would domoge piping associated with its own cooling loop and could not impact any of the isolation valves.
The electrical cabling to the pump motors is routed in conduit below the pumps.
The physical separation of these conduits was investigated. It was determined i
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that although there is a small possibility of damaging the cabling of both Pumps I and 3 with a drop of either Pump I or 3, the separation precludes impacting the Pump 2 cabling. Similarly, a drop of Pump 2 would not impact the cabling of Pumps 3 and I. Therefore, at least one pump power feed will remain intact, and sufficient component cooling water will be available.
There is also service water piping located below the CCW pump room; however, based upon the large size and physical separation of the redundant piping, unacceptable damage by scabbing of the floor above was considered improbable.
Therefore, it was concluded that the CCW pump room and area below are acceptable with regard to a potential load drop of a CCW pump motor from the monorail.
2.3 Service Water intoke Structure The service water intoke structure contains three SW pumps and strainers of 100% capacity each which are cross-connected (with valve isolation) to the two redundant service water cooling loops. The heavy load operation of interest is the lifting of one of the SW pump motors up through its individual hatch. There is the potential for dropping the motor back down the hatch or dropping it approximately 7' on the cc,ncrete deck at grade elevation which may cause scabbing of the concrete into the volve area. Therefore, the system piping, volving, and electrical feeds were investigated to determine whether sufficient physical separation and isolation capability exists and then to show that the system function would not be lost in the event of such a drop. The equipment layout is shown in Figure I1. The three SW pumps are located in the SW intake structure at elevation 578'. Each main discharge pipe runs westerly into the volve room. The pipes are separated by approximately 14'. A drop of a pump motor back down its hatch would not impact the other pumps or piping, due to odequate physical separation. A drop of the pump motor on the concrete at elevation 600', which is immediately above the pumps, may cause scabbing; however, as the lifts are confined to an area which is not over the other pumps, this scabbing would not be sufficient to disable both other pumps and piping.
Therefore, it is concluded that a pump motor drop would not disable the other SW pumps or piping inside the pump room.
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1 The service water cross-connect headers and volving are located in the valve room odjacent (to the west) to the pump room. The three pump discharge pipes are cross connected with dual isolation volves to feed the redundant loop pipes.
These pipes proceed westerly to the SW tunnel and are.<eparated by 16' within the room. in addition, the SW redundant loop piping is cross-connected with dual isolation valves, in each loop, to form a single feed to certain nonredundant components. Because the only potential load drop damage would be due to localized s'cabbing of the roof above, it was determined that both loops would not be damaged simultaneously, and that any damage could be isolated in a manner to keep at least one SW loop in operation.
It is recommended that movement of a pump motor be restricted to keep it over its corresponding loop.
The electrical cabling to the pump motors is routed in conduit to the three separate service water pumps. The pump power cabling feeds in from the west through the volve room and into the pump room to the pumps. The feeds are separated from each other by greater than 10' in the valve and pump room.
Based on this physical separation of the power feeds, it was concluded that a load drop of one pump motor would not disable both other pump power feeds in the pump room and scabbing of the valve room roof would not disable both other pump power feeds.
A number of motor control centers (MCC) are located along the SW pump room east wall. A SW pump drop could impact one or two of the MCCS; however, the equipment supplied from the MCC(s) is either not critical to the SW system operation or redundant to equipment fed from other MCCS which would not be simultaneously impacted.
l A-17
TABLEI SYSTEMS EVALUATIONS OF REGION I CASES I A, IB, IC,2 and 3 IS SYSTEM SYSTEMS OF EQUIPMENT IN REGION CONCLUDED INTEREST POTENTIALLY LOST TO BE LOST?
REMARKS RCS Vessel nozzles (hot and No Note I cold legs)
ECCS Core flood nozzles No Note 2 Notes 1.
Although hot or cold leg nozzles are potentially damaged, reactor vessel integrity is not lost. This conclusion is based on structural evaluation of the missile shield drop and reactor head drop.
2.
Rupture of one core ficod nozzle was assumed. The other nozzle was assumed ovalloble.
l
TABLE 2 SYSTEMS EVALUATIONS OF REGION 2 CASES IB, IC,2 and 3 IS SYSTEM SYSTEMS OF EQUIPMENT IN REGION CONCLUDED INTEREST POTENTIALLY LOST TO BE LOST?
REMARKS RCS Pzr. Spray Line No Note i Pzr. PORV ECCS Core Flood Tonk l-2 No Note 2 ECCS LPl/DHR thru Core Flood No Note 3 Tonk 1-2 Injection Line Instrumentation
- Pressurizer Pressurizer Level No Note 4
- Steam Generator Steam Generator Level No Note S Notes I.
The Pressurizer Spray, and PORV lines are located in this region. If these lines were ruptured, on RCS break would occur. Domoge would be minimal, even if the RCS were pressurized, because of the break location.
2.
Core flood tanks are not needed in this scenario.
3.
LPl/DHR injection through the other core flood line (Tonk l-l) would be available.
4.
All pressurizer level indication transmitters are located on the north side of the east D-ring at elevation 585'. The circuits for two transmitters are routed to the east, and the other is routed to the west.
5.
It would be expected that, at most, level instrumentation for one steam generator would be lost.
The other steam generator (including instrumentation) would be available.
TABLE 3 SYSTEMS EVALUATIONS OF REGION 3E CASES I A, IB, IC,2 and 3 IS SYSTEM SYSTEMS OF EQUIPMENT IN REGION CONCLUDED INTEREST POTENTIALLY LOST TO BE LOST?
REMARKS RCS Entire "A" Loop, including steam No Note I generator 1-2, RCPs 1-2-1 and I-2-2, pressurizer, cold / hot leg 1
piping, Pzr. surge and spray line piping, Pzr. code safety volves (2)
MU&P Normal makeup, through HPI Yes Note 2 Injection Line (Cose 2)
ECCS DHR Suction Line No Note 2 2 HPI Lines Secondary System AFW Supply to "A" Steam Yes Note 3 Generator Instrumentation
- Pressurizer Pressurizer level Yes Note 4
- Steam Generator SG Level No Note 5
- Reactor Coolant RC Wide Range (Cold Leg)
No Note 5 Temperature Notes I.
A number of scenarios were considered, including various ruptures of the RCS and crimping of lines. A scenario combining RCS break concurrent with crimping of both the pressurizer surge line and the DHR suction line was not considered plausible.
2.
DHR suction was analyzed as ruptured or crimped. A scenario combining a small RCS break concurrent with DHR crimping and PORV discharge line crimping was not considered plausible. The HPl piping in the "B" loop could be used for makeup.
3.
Although AFW supply to the "B" Steam Generator would be available, the AFW pumps are steam driven and will not be available.
4.
If the pressurizer were impacted all level indication could be lost. -
5.
Steam generator level indication would be available for SG l-1.
RC temperature would be available for Loop B.
TABLE 4 SYSTEMS EVALUATIONS OF REGION 3W CASES I A, IB, IC,2 and 3 IS SYSTEM SYSTEMS OF EQUIPMENT IN REGION.
CONCLUDED INTEREST POTENTIALLY LOST TO BE LOST?
REMARKS RCS Entire "B" Loop, Steam No Note I Generator 1-1, RCPs 1-1-1 and 1-1-2 and Cold / Hot Leg piping ECCS Core flood line from Tonk l-1 No Note 2 LPl/DHR thru core flood tank No Note 3 1-1 injection line 2 HPl lines No Secondary System AFW supply to "B" Steam Yes Note 4 Generator Instrumentation
- Steam Generator Steam Generator level No Note S
- Reactor Coolant RC temperature No Note 5 Notes 1.
Coses for both RCS breaks and no RCS break were analyzed.
2.
Core flood line from Tonk l-l is assumed lost; however, the other core flood tank line is available.
3.
LPl/DHR injection through the other core flood line (Tonk 1-2) would be available.
4.
Although AFW supply to the "A" Steam Generator would be available, the AFW pumps are steam driven and will not be available.
5.
SG level indication would be available for SG l-2. RC temperature indication would be ovallable for Loop A.
1 l
1
TABLE 5 SYSTEMS EVALUATIONS OF REGION 4 CASES I A, IB, IC,2 and 3 IS SYSTEM SYSTEMS OF EQUIPMENT IN REGION CONCLUDED INTEREST POTENTIALLY LOST TO BE LOST?
REMARKS RCS Loop "A" or "B" No Note 1 ECCS Core flood injection lines, No Note 2 including check volves Refueling Canal Possible perforation of No Note 3 canal liner and base, resulting in flooding of ECCS equipment Notes 1.
It was assumed that a load drop could impact only one loop, and not both simulaneously.
2.
It is improbable that both core flood lines would be lost due to o single drop.
3.
ECCS onalyses indicate that the ECCS equipment is not adversely offected by flooding of 404,000 gallons. This drop would result in about 388,000 gallons.
O T
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TABLE 6 SYSTEMS EVALUATIONS OF REGION 5 CASES N/A - FLOODING CONSIDERATION OM_Y IS SYSTEM SYSTEMS OF EQUIPMENT IN REGION CONCLUDED INTEREST POTENTIALLY LOST TO BE LOST?
REMARKS Refueling Canal Possible perforation of canal No Note I liner and base, resulting in flooding of ECCS equipment Notes 1.
If the canal drained, there would be about 355,000 gallons, which would be less than that assumed for a LOCA.
l l
1
TABLE 7 SYSTEMS EVALUATIONS OF REGION 6E CASES 2 and 3 IS SYSTEM SYSTEMS OF EQUIPMENT IN REGION CONCLUDED INTEREST POTENTIALLY LOST TO BE LOST?
REMARKS None None No Note I l
Instrumentation l
I
- Pressurizer Pressurizer Level No Note 2
- Steam Generator SG Level No Note 3
- Reactor Coolant RC Temperature - Cold No Note 4 Leg Wide Range f
Notes l.
No DHR or RCS equipment is located in this region.
2.
Pressurizer level would be available from the channel routed to the west portion of containment.
3.
SG level indication would be available for SG l-l.
4.
RC Temperature indication would be available from Loop 1.
l 1
1 i
TABLE 8 SYSTEMS EVALUATIONS OF REGION 6W CASES 2 and 3 IS SYSTEM SYSTEMS OF EQUIPMENT IN REGION CONCLUDED INTEREST POTENTIALLY LOST TO BE LOST?
REMARKS None None No Note i Instrumentation
- Steam Generator Steam Generator Level No Note 2
- Reactor Coolant RC Temperature - Cold No Note 3 Leg Wide Range Notes 1.
No DHR or RCS equipment is located in this region.
2.
Steam Generator level indication would be available for SG l-2.
4.
RC Temperature indication would be available from Loop 2.
e
TABLE 9 SYSTEMS EVALUATIONS OF REGION 7 CASES I A, IB, IC,2 and 3 IS SYSTEM SYSTEMS OF EQUIPMENT IN REGION CONCLUDED INTEREST POTENTIALLY LOST TO BE LOST?
REMARK 5 RCS Pressure DHR Suction Piping Yes Note l Boundary /DHR Suction ECCS 2 HPI Lines No Note 2 LPl/DHR Injection No Note 3 MU&P MU through one HPI Line Yes Note 2 Pressurizer DHR Alternative Pressurizer Yes Note 4 Spray from DHR Pump l-2 PORV No Note S Instrumentation
- Pressurizer Pressurizer Level No Note 6
- Steam Generator Steam Generator Level No Note 7
- Reactor Coolant RC Temperature - Cold No Note 8 Leg Wide Range RC Pressure No Note 8 Reactor Core Thermocouples No Note 9 Notes 1.
DHR suction could be ruptured or crimped.
2.
HPI would be available to other RCS Loop. It could be used for makeup.
3.
LPi would be available to other RCS Loop.
4.
The DHR system provides alternative pressurizer spray when the reactor coolant pumps are not ovallable (as in this operating condition).
S.
Although the PORV circuit is routed in Region 7, it is only in the extreme northeast portion.
6.
Pressurizer level information would be available from the circuit routed to the west in the containment.
l
TABLE 9 (continued)
Notes 7.
Steam Generator level indication would be available for SG l-l.
8.
RC temperature and pressure indication would be available from Loop B.
9.
Channel B T/Cs are routed through Region 7; however, Channel A T/Cs would be available.
t 6
r y
TABLE 10 SYSTEMS EVALUATIONS OF REGION O CASES 1 A, IB,2 and 3 IS SYSTEM SYSTEMS OF EQUIPMENT IN REGION CONCLUDED INTEREST POTENTIALLY LOST TO BE LOST?
REMARKS RCS In-Core Instrument Tube (s)
Yes Note !
ECCS LPI System, including No Sump Recirculation DHR Suction Piping Yes Note 2 HPI Lines No MU&P MU through one HPl Line Yes Note 3 Instrumentation
- Reactor In-Core Thermocouple -
Yes Note 4 Core Temperature Refueling Refueling Transfer Tubes Yes Note S System
, Notes 1.
It was assumed that in-core instrument tubes could be severed, and on RCS break could result.
2.
Loss of the DHR suction will require that cooling be performed by natural circulation; however, MU&P would possibly be lost (see Note 4).
3.
MU&P would be lost, but HPI could be used for makeup. FW would be required for SG heat removal.
4.
Loss of the instrument tank and tubes would result in loss of core T/C information. Loop temperature information would still be available, as would pressurizer level.
5.
Loss of the transfer tubes would not occur concurrent to loss of DHR.
Therefore, on alternative is, in reality, still available.
i r
TABLEII SYSTEMS EVALUATIONS OF REGION 9 CASES I A, IB, IC,2 and 3 IS SYSTEM SYSTEMS OF EQUIPMENT IN REGION CONCLUDED INTEREST POTENTIALLY LOST TO BE LOST?
REMARKS ECCS Core Flood Tonk l-1 No Note I LPl/DHR through Core Flood No Note 2 Tonk 1-1 injection Line 2 HPl Lines No MU&P MU&P Suction Line, Yes Note 3 including 2 Coolers Instrumentation
- Reactor Coolant RC Pressure No Note 4
- Reactor Core Thermocouples No Note 5 Notes 1.
Core flood is not required.
2.
Other LPl/DHR injection would be available.
3.
MU&P not required because DHR not lost.
4.
RC pressure indication for Loop A would be available.
5.
Channel A T/Cs are routed through this region; however, Channel B T/Cs would be available.
l
I TABLE 12 REGION 1 -EVENT TREE ASSESSMENT CASE PATH CONCLUSION Case I A I or 4 OK Case IB I
OK Case IC 1
OK Case 2 i
OK Case 3 I
OK i
TABLE 13 REGION 2 - EVENT TREE ASSESSMENT CASE PATH CONCLUSION Case l A N/A Case IB 3
Consider Alternative Cooling Modes (DHR)*
Case IC l
OK Case 2 i
OK Case 3 I
OK DHR would effectively be available, because the predicted RCS break would be in the high point of the system. Some makeup may be needed.
e
N TABLE 14 REGION 3E - EVENT TREE ASSESSMENT CASE PATH CONCLUSION i
Case I A I or 4 OK Case IB l
OK Case IC l
OK Case 2 19 OK Case 3 2 or 5 OK 1
t e
r
3 TABLE 15 REGION 3W - EVENT TREE ASSESSMEN f CASE PATH CONCLUSION t
Case I A I or 4 OK Case IB l
OK Case IC l
OK Case 2 i
OK Case 3 I
OK I
d e
1
TABLE 16 REGION 4 - EVENT TREE ASSESSMENT i
4 CASE PATH CONCLUSION 4
Case I A I or 4 OK Case IB I
OK i
Case IC 1
OK Case 2 i
OK Case 3 I
OK i
1 4
4 4
1 4
4 i
'a i
4 I
i
~
4 2
4 5
h J
,K I '/
( \\
14 t.
j t
6 i
1 TABLE 17
. t l
\\
J
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REGION 5 - EVENT TREE ASSESSMFAT t
,N CASE PATH CGNCLUSION i :
Case I A N/A O
Case IB N/A 4
4 Case IC N/A Case 2 i
OK 4
Case 3 I
OK
,s-3 1
i 4
}
i t
l
,t
-l l
l t
i l
5 s
s 4
6 k
l 1
e 4
i s
k i
e I
'l
,.>a=~
ll 5
,.. l n l1
TABLE 18 REGION 6E - EVENT TREE ASSESSMENT CASE PATH CONCLUSION
(
Case l A N/A Case IB N/A Case IC N/A Case 2 i
OK Case 3 I
OK l
E
TABLE 19 REGION 6W - EVENT TREE ASSESSMENT CASE PATH CONCLUSION Case l A N/A L
Case IB N/A Case IC N/A Case 2 i
OK Case 3 i
OK 4
I
- ~ ~ - -,,
1 i.
TABLE 20 1
REGION 7 - EVENT TREE ASSESSMENT i,
)
i i
CASE PATH CONCLUSION i
l Case l A I or 4 OK i
1 Case IB l
OK Case IC 1
OK 1
I Case 2 19 OK Case 3 2
OK 1
a I
i 4
ti i
Y 1
l i
1 I
I.
4 s
I-l l
I l
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~
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p q
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,--nwwe<+++
e-
~-
wer-~
--~
--v, w vr ea
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.-o w,,~,.,-v-er
TABLE 21 REGION 8 - EVENT TREE ASSESSMENT CASE PATH CONCLUSION Case I A I or 4 OK Case IB 1
OK Case IC N/A OK Case 2 2 or 19 OK Case 3 3
Consider Alternative Cooling Modes (DHR)*
Based on the separation of DHR and the fuel transfer tubes within this region, one or the other is considered to be available.
- = -
y.,
m.
1 TABLE 22 REGION 9 - EVENT TREE ASSESSMENT 4
CASE PATH CONCLUSION Case I A l
OK Case IB l
OK
. Case IC i
OK t
Case 2 i
OK Case 3 i
OK l
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4 FKAREI LOAD DROP EVALUATION REGION I -REACTOR VESSEL 3
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a
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- .. :W:
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e FIGURE 2 LOAD DROP EVALUATION REGION 2 - NORTH EPO OF OPERATING DECK - 603' EL l
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4 FIGURE 3 LOAD DROP EVALUATION REGION 3E Ato 3W - D. RING ENCLOSURES l
l I
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- %~
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S FIGURE 4 LOAD DROP EVALUATION REGION 4 -REFUELING CANAL
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FIGURE 5 LOAD DROP EVALUATION REGION 5 - SOUTH EPO OF REFUELING CANAL l
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i FIGL#tE 7 LOAD DROP EVALUATION REGION 7 - GRATNG N 5OUTEAST QUADRANT - M EL l
l i
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____. s%afh i
(f.%
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I 9
FIGURE 8 LOAD DROP EVALUATION REGION 8 - H. CORE INSTRUMENT AREA
g 4
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? Aaf%
A JcM L> !
I se s e
aee l
FIG.RE 9 LOAD DROP EVALUATION REGK)N 9 - AREA ADJACENT TO EOUIPMENT HATCH - SW QUADRANT - 607 EL
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i i
PRIMARY COOLING MODE BWST LPI OR HPI OR MU RECIRC (SUMP & LPI)
I 1.
OK 2.
3.
4.
OK REFUELING CANAL FILLED; DRAINS TO SUMP THRU RCSBREAK 5.
6.
Consider Alternative Cooling Modes FIGURE 12 CASE I A - RV HEAD REMOVED - RCS BREAK
COOLING MODE BWST LPI OR HPl OR MU PORV (SU PI) l.
OK 2.
3.
4.
5.
I t
Consider Alternative Cooling Modes
(
l FIGURE 13 CASE IB - RV HEAD IN PLACE - SMALL RCS BREAK 1
COOLING MODE BWST LPI OR HPI OR MU RECIRC (SUMP & LPI) i 1.
OK i
2.
i
'i 1
3.
4.
f l
i i
l l
1 l
Consider Alternative Cooling Modes l
FIGURE 14 CASE IC - RV HEAD IN PLACE - LARGE RCS BREAK
PRIMARY COOLING BACKUP COOLING MODE I BACKUP COOLING MODE 2 MODE MU&P STARTUP DHR PUMP BWST AND RECIRC DHR OR HPI PZR.
HPI OR LPI PORV (SUMP OR LPI l l
l PUMP l SPRAY)
OK 2.
OK 3.
OK-4.
5.
6.
7.
OK 8.
9.
- 10. *
- 11. OK
- 12. *
- 13. *
- 14. *
- 15. OK
- 16. *
- 17. *
- 18. *
- 19. OK
- 20. *
- 21. *
- 22.
- Consider Alternative Cooling Modes FIGURE 15
)
CASE 2 - RV HEAD IN PLACE - NO RCS BREAK
PRIMARY COOLING BACKUP COOLING MODE I MODE FLOW THRU FUEL XFR.
DHR BWST LPI p g sb d5bt s)p T
T COOLING BY SFPCS
- 1. OK
- 2. OK 3.
4.
(REFUELING CANAL FILLED) 6.
Consider Alternative Cooling Modes FIGURE 16 CASE 3 - RV HEAD REMOVED - NO RCS BREAK
APPENDIX B STRUCTURAL EVALUATIONS 1.0 STRUCTURAL EVALUATION INTRODUCTION Structural load drop evaluations are initiated at two levels:
l.
As early input to the systems evaluation to identify drop scenarios that potentially lead to discrete levels of domoge (e.g., local concrete scabbing versus gross struc-tural failure) and; 2.
Following bounding systems, dose, and criticality evalua-tions when these initial approaches fail to demonstrate acceptable consequences.
The effort associated with item I is qualitative, yet based upon a wealth of quantitative information and experience, while that for item 2 is fully quantita-tive and plant specific.
Upon the identification of all heavy loods (including a full characterization of the weights, dimensions, material properties, and structural chorocteristics) and the locations where they are handled (reference Tables 1-5), load drop scenarios based upon realistic consideration of plant procedures are evaluated to identify loads which control:
l 1.
local response (e.g., penetration, scabbing, spalling, per-foration, etc);
2.
overall structural response (e.g., large inelostic deforma-tions or obrupt failures of principal structural members, etc), and/or; 3.
loads that may induce behavior that exhibits combined response such that either overall or local failure modes would control.
The results of this evaluation are tabulated in Table 6.
This information is factored directly into early systems evoluotions.
B-1
Local effects are generally independent of the dynamic chorocteristics of the impacted structure; however, overall structural response is a direct function.
Where the controlling mode of response is listed as " local," these loads are considered copoble of slob perforation or scabbing of the concrete deck backfoce. It was found that many postulated drops are copoble of producing scobbing; therefore, it was decided that the consequences of scabbing must be considered in the systems evoluotion.
That is, all systems which could be impacted by the scabbing were assumed to be lost.
Postulated drops of the Automatic Reactor inspection System (ARIS), the polar crane load block, the reactor vessel head lifting rig, and various hatch covers fall in this category and bound other load drops that potentially lead to local effects.
A discussion of local effects evoluotion techniques follows in Section 2.1.
Where the controlling mode of response is listed as "overall structural," these foods are considered capable of producing gross and intolerable distortions of primary structural members and possibly propogoting failures. Postulated drops of the reactor missile shields, the reactor vessel head, and reactor plenum ossembly fall in this category and bound other load drops that potentially lead to "overall structural" effects. A discussion of overall structural response evalua-tion techniques follows in Section 2.2.
Detailed structural evoluotions are initiated when it becomes apparent that the consequences of bounding systems, dose, and criticality evoluotions are unoc-ceptable. These evoluotions are intended to establish a more realistic ossess-ment of the domoge potential of specific load drop scenarios. The information gained from these evoluotion is fed back into the systems evoluotions at a later stage. The results of these specific structural load drop evoluotions follows in Section 3.0.
2.0 STRUCTURAL EVALUATION METHODOLOGY AND CRITERIA The evoluotion methodology and criterio generally follow the recommendations mode by the American Society of Civil Engineers Technical Committee on B-2
Impulse and Impact Loads (Reference 7). These recorr.mendations are supple-mented 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 evaluation methodology and criteria which are addressed in Sections 2.1 and 2.2 consider the two potential modes of structural behavior, local effects and overall structural response, respectively.
2.1 Local Impact Response Evoluotion Local impact response may lead to severe damage such as crushing, perforation, and concrete ejection in the vicinity of the impactive load; however, overall dynamic response of the structure in the form of reactions away from the load are insignificant. The complex nature.of local impact response of reinforced concrete requires evoluotion using empirical formulae that are experimentally derived. The modified National Defense Research Committee (NDRC) formula (Reference 8) was chosen because it has been shown to give the best fit with available experimental data (Referen4es 9 and 10). The NDRC formulae for the depth of penetration, x (inches), of a solid cylindrical missile are given by:
)1.81/2 for 5 s 2.0 (I) 4 KNWd ( V d
x=
1000d or l.8 x=KNW V
+d for 5 22.0 (2) 1000d d
t 1
where W = weight of the missile (pounds) d = diameter of missile (inches)
V = impact velocity of missile (feet /second)
N = missile shape factor
= 0.72 flat-nosed missiles
= 0.84 blunt-nosed missiles B-3 J
en
= 1.00 spherical-nosed missiles
= 1.14 sharp-nosed missiles K = concrete penetrobility factor
= 180/5(f'c = concrete compressive strength in pounds / square inch)
The thickness of reinforced concrete needed to resist impact without perforation and scobbing are given by the following Army Corps of Engineers formulae which can be used in conjunction with equations I and 2 (Reference il).
ts = 2.12 + 1.36 ({ for 0.65 < f < lI.75 (3) d (d /5) for 1.35 < 5 < l3.5 (4) tg = 1.32 + 1.24 d
d where is = concrete thickness required to prevent scabbing tp = concrete thickness required to prevent perforation E,quotions 3 and 4 were later extrapolated for small values of x/d (Reference 12)
- giving, t = 7,9l { 3 ) - 5.06 (
for * < 0.65 (5) 2 g = 3.19 ( * ) - 0.718 ] }
for * < l.35 (6) t A 10 percent margin on thickness has been applied in the use of equations 3 through 6 as recommended in Reference 7, except for concrete sections backed by steel decking where the equations were used directly.
The effects of shape and deformobility have been conservatively accounted for in the case of the NDRC formula by adjusting the missile shape factor, N, and/or using " equivalent" diameters.
B-4
2.2 Overall Structural Response Evoluotion Overall structural response results from the dynamic interaction of the impact-ive load 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 load. As a rule, this forcing function is unknown; however, occasionally it con be estimated by incorporating knowledge of the chorocteristics of the dropped load (weight, size, shape, deformability), characteristics of the impacted structure (material properties, structural configuration), and the impact conditions (velo-city, orientation).
The following discussions address the use of energy balance methods for the evoluotion of reinforced concrete and structural steel structures.
These techniques do not require explicit knowledge of the forcing function.
Reinforced Concrete Structures The load drop methodology shown in Figure I 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 absorbed by the concrete floor system without reaching those limits of unocceptable behavior is balanced against the externally applied energy resulting from a heavy load drop.
It has been assumed that momentum is conserved, and the kinetic energy of the drop drives the mass of the floor and induces strain.
As on additional conservatism, no credit has been taken for potential sources of energy dissipa-tion through local deformation in the forms of concrete crushing and penetro-tion.
B-5
Generally, the ultimate load of a concrete slab or beam system is reached prior to exceeding the hinge rotational capacity of particular sections provided that on unstable mechanism has not formed. The hinge rotational capacity was used as a criterion to set the maximum allowable level of deflection for the concrete slob or beam system. The hinge rotational capacity for concrete structures was developed in References 13 and 14 based on test results given in References 15 ond 16 and is given as:
ru = 0.0065 (d/c) 5 0.07 (7) where r = rotational capacity of plastic hinge (radians) u 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 a concrete slob or beam with a plastic hinge at its center is then given by:
Xm = (r l/4)
(8) u
- where, Xm = mcximum deflection L = span of beam Rotations of the magnitude governed by equation 7 result in cracking which is confined to a region below (above) the tensile reinforcement.
Generally speaking, the section will remain intact with no crushing, spalling, or scabbing due to flexure; however, scabbing may occur as a result of shock wave motion associated with the reflection of tensile waves from the rear surface or shear plug formation. It has been conservatively assumed that scabbing does occur.
l l
l B-6
~
Tha lond/ deflection history up to the point of the ultimate loading, coupled with the maximum allowable deflection, defines the maximum level of strain energy obsorption provided that a shear failure has not occurred. The shear stress at limiting sections was checked and compared to allowables os specified in Chapter 11 of ACI 381-77 (Reference 17).
Structural Steel Structures The maximum response of structural steel elements !s determired using the commonly applied energy balonce method (References 7,18, and 19) 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 as:
(9)
,,Um 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 7 and are es follows:
ALLOWABLE MODE OF RESPONSE DUCTILITY RATIO
- l. Flexure
- open secticos 12.5
- closed sections 20
- 2. Shear 5
- 3. Compression 14 x 104 < 10 F (K_L)2 y r 0.5
',"7
- 4. Torsion B-7
where Fy = minimum yield stress of the steel K = theoretical effective length factor for compression member L = length of compression member cu = ultimate strain Ey = yield strain The effective yield limit corresponds to the inflection point of on equivalent elasto-plastic resistance displacement curve os defined by Newmark (Reference
- 20) and shown in Figure 2. 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 hordening mechanism.
2.3 Discussion of Structural Margins in addition to the conservatisms previously mentioned, the following conserva-tisms are also inherent in the methodology used in the evoluotion:
1.
Static material strengths for concrete and steel are used.
Test data shows that this property increases with the increased strain rates associated with dynamic loadings.
For excmple, References 19 and 21 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.
2.
Design (minimum) material properties for concrete and steel are used. No increase is taken for the aging of concrete which con amount to a factor of up to 1.35 (Reference 22) of increased strength. Also, the overage strength for structural steel is nearly a factor of 1.25 (Reference 23) higher than the minimum yield require-ment specified by ASTM.
While these factors above minimum ccJo strenath exist and contribute to structural margins, they are not used in the evoluotion.
3.
Equation 7 for hinge rotational capacity is used. This corresponds to rotations of the order of 2 degrees with minimum cracking and no crushing or scabbing. To meet necessary performance requirements (i.e., holting propo-goting failures), larger rotations in the range of 5 to 12 degrees could be tolerated. Such rotations would leod to crushing, spalling, and scabbing of the section (Reference 21); however, overall load carrying capability is expected B-8
to remain intact. Experimental observations (Reference
- 24) suggest even further capability for well-designed and well-onchored slabs. Failure 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 slab. Use of these larger rotational capabilities would have resulted in greater energy obsorbing capabilities of the floor system.
4.
The onolysis uses ACI 318-77 allowable shear stresses. A significant body of dato suggests the existence of higher shear copobilities on the order of 10[c to 20[c (References 25 through 33).
5.
The structural loads are distributed directly under the dropped heavy lood. In reality, a more favorable load distribution would exist due to the load distribution capa-bility of the slab.
6.
No credit is token for local energy dissipotion associated with any crushing of the load itself or the immediate surface of the floor.
3.0 STRUCTURAL LOAD DROP EVALUATIONS INSIDE CONTAINMENT -
RESULTS, CONCLUSIONS, RECOMMENDATIONS The following sections address specific structural load drop evaluations for postulated drops within each region inside containment.
These evaluations represent those which are considered to be bounding in terms of core cooling and radiological consquences.
3.1 Drops Within Region I 3.1.1 Plenum Assembly Drop Onto the Reactor Core The plenum osssembly, located directly above the reactor core, is removed as a single component before refueling. It weighs approximately l19,000 pounds with its lifting rig and is removed and replaced according to System Procedure l
SP 1505.1, Reactor Internals Removal and Replacement (Reference 34). The lifting system used to move the plenum includes the containment polar crane, the plenum assembly lifting rig, and various adapters, pendants, and fixtures.
These items have been evoluoted (Reference 1) and found to meet the intent of l
B-9 i
industry standard ANSI Bl4.6-1978 (Reference 35) with a corresponding factor of safety against reaching ultimate strength of at least five.
Reference I o!so proposed a new procedure, " Periodic Test Procedure for Special Lif ting Devices,"
which is being implemented to provide increased assurance that these devices would be able to maintain their design load margins.
The polar crane was evaluated to industry standards CMAA 70-1975 (Reference
- 36) and ANSI B30.2-1976 (Reference 37) and found to meet these standards with two minor exceptions. Justification for these exceptions has been provided in Reference 2.
Notwithstanding the fact that the lif ting system, including the containment polar crane and associated lifting devices, complies with the intent of applicable industry standards and possesses demonstrated margins to failure, on evoluotion has been performed for a postulated drop of the plenum onto the reactor core from the highest possible elevation as dictated by physical constraints.
The plenum assembly consists of a plenum cover, upper grid, control rod assembly (CRA), guide tube ossemblies, and a flanged plenum cylinder with openings for reactor cooloni outlet flow. The plenum cylinder (127-3/4" height) consists of a large cylindrical section with flanges (upper flange has 166-7/8" O.D.) on both ends to connect the cylinder to the plenum cover and the upper grid.
The plenum cover is constructed of a series of parallel flat plates intersecting to form square lattices;it has a perforated top plate and on integral flange at its periphery. The cover assembly is attached to the plenum cylinder top flange. The perforated top plate has matching holes to position the upper end of the CRA tubes. Lifting lugs are provided for remote handling of the plenum assembly. These lifting lugs are welded to the cover grid. The upper grid consists of a perforated plate which locates the lower end of the individual CRA guide tube assembly relative to the upper end of a corresponding fuel assembly. The CRA guide assemblies provide structural attachment of the grid assembly to the plenum cover by welding to the plenum cover top plate and bolting to the upper grid.
B-10
Locating keyways in the plenum assembly cover flange engage the reactor vessel flange locating keys to align the plenum assembly with the reactor vessel, the reactor closure head CRD penetrations, and the core support assembly. The bottom of the plenum assembly is guided by the inside surface of the lower flonge of the core support shield.
During removal and replacement of the plenum, alignment is accomplished using the internals indexing fixture which is positioned on the reactor vessel flange.
The internals indexing fixture is cylindrically shaped with internal locating keys that mate with the locating keyways in the plenum assembly cover flange.
Perfect alignment of the plenum within the internals indexing fixture is required to slide the plenum in out of the reactor vessel; therefore, the maximum postulated drop height corresponds to the indexing fixture height which is 73Y2 inches.
Conservatively neglecting the energy obsorbing effects of drag as the plenum travels through water and a "doshpot" effect that exists due to the close tolerance of the plenum within the core borrel, the kinetic energy of the drop is 729,000 foot-pounds.
The total kinetic energy has conservatively been assumed to reach the core and directly load the fuel assemblies, although certain physical limitations exist and potential odditional sources of energy disspitation have not been fully quantified.
The impoet load is transmitted uniformly from the plenum upper grid to the fuel assembly upper end fittings through the 16 control rod guide tubes, and to the fuel assembly lower end fittings. The fuel rods are not significantly loaded unless the upper end fittings are driven into the fuel rods due to deformation of the guide tubes through buckling. The energy absorbed by the guide tubes failing in on inelostic buckling mode has been conservatively ignored.
Individual fuel rods are predicted to buckle elastically between spacer grids at a Euler critical buckling load (Per) of 240 pounds. Strain energy can be obsorbed beyond the point of reaching Per through bending until the fuel clodding strain reaches a value of I percent. This strain criterion is based upon the irradiated properties of Zircollory-4, the clodding material.
B-lI
The fuel rod is assumed to take a half wave sinusoidal shape between spacer grids based upon a pinned-pinned boundary condition (see Figure 3). Accordingly, the deflection along the fuel rod is given by, Y=Asinf (10) where f = length of fuel rod between spacer grids A = lateral deflection of fuel rod at mid span X = distance along span Y = lateral deflection of fuel rod at a distance x olong the span From beam theory, f=
E AI sinf (ll) where R = radius of curvature. At midspan, 2
h= A(f)
(12)
Also, assuming plane sections remain plane (see Figure 4).
K, c *t 1
(13)
D where
<c
= strain at extreme compression fibre
't
= strain at extreme tension fibre D = diameter of fuel rod i
e i
B-12
From Figure 4, q=c l-1 (14) and c
- c-*T "
(15) where e " = yield stroin.
y Combining (13) and (15);
1_*1 (16)
R cD and combining (12) and (16) 2 c A = (f) d (17)
From Figure 4, 1
)
f I\\ Y.- 9 l
(18)
=1-2q cos $1 - 2D D
(19) cos $2 = 1 - 2q - 2c cos $3 = 1 - 2q - 4c (20)
- Letting,
=
$4=w-$3 (21)
B-13
From the stress distribution shown in Figure 4,
'l
'3 cos 4 - cos *2 P = 2 / Fytrdt + 2 / Fy trdt (22) cos 4 - cos 4 1
2 44
$3 where P = oxial load t = cladding thickness r = radius of fuel rod to center of cladding F = yield stress.
y Evoluoting the integrals, P = 2Fytr(4 -4 ) + 2Fytr(sin 4 -sin $ -cos 4 (# ~*1)/(cos 4 -cos 4 ))
3 4 3
1 2 3 3
2 (23)
Similarly, the moment is given by, 4
s(p[rtrdt M=2 / Fy (cos 4 - cos 4 )rtrdt + 2 0
2
/ Fy cos,3-cos 42 (24)
Evoluoting flw integrals, 2
M = 2Fytr sin 43 + sin 43 - cos $ I'l~'4)
+
2
( +
5 2)I'3~'1)+
sin 2 43 - sin 2 4 )
cos 43 cos 42 1
-2cos$(sin 43 ~ SI" 4 )
(25) 2 1
\\
i B-14 l
Expressions (23) and (25) define the axiol force and moments in the fuel rod as a function of the changing inelastic stress distribution. Together these expressions define the midspan deflection A, given as A=M (26)
P Referring to Figure 3, the shortening of the fuel rod is given by, a=A2,2 (27) 4,,
From (26) and (27), the post-buckling lood-deflection diagram (see Figure 5) for a fuel rod can be drown. The crea under this curve represents the strain energy obsorbed by the fuel rod.
The strain energy required to be absorbed to balance the externally applied kinetic energy of 729,000 foot-pounds corresponds to on oxial deflection of 0.180 inches per span (see Figure 5). At this point, 55 percent of the fuel rod fibres measured along the diameter have reached the yield stress. The strain in the extreme compression and tension fibres is approximately 0.0074 and 0.0054, respectively. These values are less than the acceptance strain of 0.01.
Based upon our evoluotion, in the unlikely event that the polar crone and its associated lif ting devices fail while the plenum is at the maximum point of carry at which it con impact the core, we conclude that the fuel cladding will not rupture and radioactive gases will not be released.
3.1.2 Reactor Vessel Head Drop Onto the Reactor Vessel The reoctor pressure vessel (RPV) head is hemispherically shaped and weighs opproximately 330,000 pounds with the control rod drive (CRD) service structure and the RPV head lif ting rig. The RPV head has on I.D. = 165 inches'and a flange O.D. = 210 inches, giving a flonge crea of opproximately 13,250 square inches.
B-15
The RPV head is removed and replaced according to System Procedure SP 1504.01, Reactor Vessel Closure Head Removal and Replacement (Reference 38). The head lifting system is essentially identical to the plenum assembly lif ting system, with the exception of different pendants. Similarly, the lifting system complies with the Intent of industry standards as previously described.
The head is moved laterally along a direct path between the reactor vessel and the head storage stand, which is located at the north end of the operating floor.
SP 1504.01 allows the head to be removed without simultaneously raising the refueling canal water level if radiation levels permit. The polar crane main hook is used at slow speed to raise the head to elevation 578'-5", I inch above the RPV flange, at which point the hook is stopped to assure that the head is level and that the lif ting system holds the load. The head is then raised unti' the tops of the olignment studs (elevation 580'-8-7/16") are cleared.
SP 1504.01 has been revised to require that the head be moved north and away frem the RPV, holding this elevation prior to being lifted to elevation 609'-7", the elevation of the head storage stand. The head is replaced in a reverse sequence.
A postulated drop of the RPV head of 5'-0" was considered. Energy dissipation due to o transfer of momentum (References 18 and 39) was conservatively ignored. This reduction is typically of the order of 50 percent or greater for the mass ratios involved. The impact load path is from the RPV flange through the RPV shell to the cold leg nozzles from which the RPV is supported. Each RPV support consists of two heavy section double I-beam assemblages which are embedded in the concrete primary shield wall. A 44-inch section contilevers out from the wall.
An l-beam cross member spans between each of the two embedded double I-beam assemblages to support the RPV nozzle support shoe.
The limiting mode of behavior is governed by shear within the contilevered I-beams. The ultimate load of each support is approximately 7 million pounds.
Given on allowoble shear mode ductility of 5, the strain energy obsorbed before failure is 471,000 foot-pounds per support (1.88 million pounds for all four supports). The kinetic energy for the postulated drop is 1.65 million foot-pounds; therefore, significant margin exists against a shear failure of the RPV supports.
B-16
I The bending and shear stresses in each of the RPV cold leg nozzles were checked using procedures documented in Reference 40. All stresses were found to be below the yield stress level for the material. Therefore, the leak tight integrity of the reactor coolant pressure boundary is expected to remain intact.
3.l.3 Missile Shield Drop Onto the Control Rod Drive Service Structure Six concrete missile shields weighing 94,500 pounds each provide protection ogainst a postulated control rod ejection occident. The missile shield panels are approximately 30 feet long, 3 feet thick, and 6.5 feet wide. The shields are positioned at the 653 foot elevation and span between the two steam generator D-ring enclosures. Prior to reactor disassembly, the shields are moved by the polar crane and placed on either of the D-ring enclosures.
The load carrying reliability of the lifting system is such that a drop of a missile
{
shield is highly improbable. However, to increase the safety of these lifts even further, procedural changes have been made to require that the shields be removed and replaced in the following sequence:
Removal Replacement 3 and 4 l ond 6 i
2 ond 5 2 and 5 l
I and 6 3 and 4 l
where the numbers correspond to individual shields os they line up from north to south in their normal operating position at the 653' elevation. These procedures require that each of the interior shields (those closest to the RPV center line; numbers 2 through 5) be lifted vertically until they just clear their holddown studs, and then translated north (shields 2 and 3) and south (shields 4 and 5) prior i
to the east or west translation for laydown top of the D-rings.
1 These procedural changes ensure that for a postulated shield drop as it is just lifted over its normal position, the shield would fall directly back and come to rest on the D-ring ledge. For a postulated drop of shields 2 through 5 while over shields I or 6, the dropped shield is prevented from o fall. This eiiminates the B-17
potential for either o translating drop from height or o rotating drop over the D-ring edge. Drops of shields I and 6 will not impact the RPV service structure; however, they potentially could impact the refueling cavity and, therefore, this scenario was considered (see Section 3.4.1).
Based upon our evoluotion and in con.tideration of the procedural changes, we find that the consequences of a postulated missile shield drop are acceptable.
3.2 Drops Within Region 2 3.2.1 Reactor Vessel Head Drop in Vicinity of Reactor Vessel Head Storage Stand During reactor disassembly, the RPV head is stored on the reactor head storage stond which is located at the north end of containment on the elevation 603' operating floor. The head must be lifted 6'-7" to elevation 609"-7" to be set on the stond. A postulated drop from o height of 7 feet onto the floor area just south of the stand was considered.
The 8-inch concrete refueling floor in the drop region is supported by a structural steel grid. Five W36 x 135 steel beams spanning 16 feet are supported off of the refueling cavity wall on the south side and a column supported W36 x 300 steel beam on the north side. Credit was not given to the 8-inch concrete slab for either load resistance or the capability to distribute the load.
Accordingly, o model was developed where the steel grid was loaded directly at points corresponding to the contact points of the RPV head flange with the grid.
The ultimate load carrying capacity and deflections at the ultimate load were calculated for each member of the grid. The allowable ductilities specified in Section 2.2 were used to determine the energy absorbing capacity of the floor in this area. It was determined that the head could be lif ted opproximately a foot and a half over the operating floor; however, a drop from the 7-foot height needed to place the head on its stond could not be tolerated without excessive deformations. The consequences of these deformations are evoluoted in Appen-i l
dix A, Section 2.1.2 l
l t
B-18
3.2.2 Miscellaneous Drops Toble 5 was utilized to help identify heavy loads that, in addition to the reactor vessel head, may be bounding for postulated drops in other areas within Region 2.
The reactor vessel head lifting rig was chosen as a bounding load due to weight and carry height considerations.
The local impact response methodology documented in Section 2.1, Local impact Response Evoluotion, was utilized to assess the potential for scabbing or perforation of the 8-inch concrete slab of the 603' elevation. Based upon our evoluotion, we conclude that both scabbing and perforation are possible given postulated drops onto the floor. Accordingly, the consequences of scabbing and perforation have been considered in the systems evaluation (reference Appen-dix A, Section 2.1.2).
3.3 Drops Within Region 3 3.3.1 Missile Shield Drops on Top of D-Rirjg During reactor disassembly, the missile shields are stored in top of the D-ring enclosures at the 653' elevation (three on each D-ring). The shields are carried at a minimum height and moved in on east or west direction depending upon on which D-ring they are stored. The shields may be oriented either north-south or east-west when in their laydown position.
The shields are supported by heavy section structural steel members (1-beams) that span east-west between the D-ring concrete walls. Removable grating is attached to the I-beams which are also removable for major repairs such as a reactor coolant pump motor removal.
A north-south orientation is structurally bounding for a postulated shleid drop, since the food from on east-west orientation drop would go directly into the D-ring walls which have substantial structural resistance. A flat east-west orientation drop would impact seven I-beams which collectively have on energy obsorbing capability of approximately 2 million foot-pounds. A maximum drop B-19
height was calculated utilizing the methodology cnd criteria douemented in Section 2.2, Overall Structural Response Evaluation, and found to be opproxi-mately 20 feet for a flat drop. Thus, substantial margin exists for lif ts above the normal carry height of 3-4 feet as necessitated by the practice of stacking shields two high. Furthermore, since many beams are active in absorbing energy, postulated oblique drops resulting from failure of a sling on one side are also acceptable.
Based upon our evaluation, wc conclude that a postulated missile shield drop onto the D-rings at elevation 653' will not result in unacceptable structural conse-quences.
3.3.2 Miscellaneous Drops Table 5 was used to help identify heavy loads that may be bounding for postulated drops within the D-ring enclosures when the 1-beams and grating are removed for major repairs. A postulated drop of a reactor coolant pump motor was found to be bounding based upon weight and carry height considerations.
A scoping evaluation was made to assess the efficacy of conducting a structural evaluation to estimate the consequences of this postulated drop. A judgment was made not to initiate such evaluations due to the complexity and cost of the evaluation process that would be required to establish a defensible evoluotion of the integrity of the reactor coolant pressure boundary. Accordingly, the systems evaluations (reference Appendix A, Sections 2.l.3 and 2.1.4) considered a loss-of-coolant occident and various small line breaks as consequences of this postulated drop.
3.4 Drops Within Region 4 3.4.1 Missile Shield Drop Onto the Cavity Floor The missile shields represent the bounding heavy loods carried over the refueling cavity (elevation 578'). The shields are normally positioned over the cavity at the 653' elevation; therefore, a postulated drop height of 76 feet could be B-20
1 assumed, including a one foot margin. Although procedural changes limit the probability of dropping shields 2 through 5, physical means (other than the proven load carrying reliability of the lifting system) do not exist to mitigate a postulated drop of shields I or 6 (see discussion in Section 3.1.3).
Therefore, both local and overall structural response evoluotion methodologies documented in Section 2.0 were incorporated to evoluote structural consequences of a postulated end-on shield plug drop onto the cavity floor.
The refueling cavity concrete floor is 4 feet thick and lined with a stainless steel liner. The cavity spans approximatly 28 feet in the short direction. The floor slob is reinforced with #8 rebor at 6 inches, top and bottom running east-west, and #8 rebor at 8 inches, top, and #8 rebar at 6 inches, bottom, running north-south. The ultimate load resistance was calculated incorporating yield line analysis assuming a circular yield fan for a point loading. Overall structural response effects were found to be bounding over local response effects; however, unacceptable deformations were predicted.
Based upon these analyses, the systems evoluotions (reference Appendix A, Section 2.l.5) have considered the consequences of damage to piping and equipment below the cavity floor.
i 3.4.2 Plenum Assembly Drop Onto the North Cavity Floor During normal refueling, the plenum assembly is removed from the reactor following System Procedure SP 1505.01, Reactor internals Removal and Replace-ment, and is stored in the north end of the refueling cavity on a cylindrical stond that is 44h inches high. The plenum must clear this height to be placed on the stand. A structural evaluation was made to determine the overall structural response of the cavity slab for o postulated drop of the pienum in the vicinity just south of the plenum assembly storage stand.
The overall structural response evaluation methodologies and criterio document-ed in Section 2.2 were incorporated. The ultimate lood resistance of the cavity slob was determined by yield line analysis conservatively assuming a diogonal yield pottern. In view of the large diameter of the plenum and the fact that the B-21
slab would be loaded uniformly over o brood area, the load resistance and energy obsorbing capacity would be even greater than assumed.
Based upon our evaluations, we have determined that a postulated drop of the plenum assembly onto the cavity floor will not lead to unacceptable deformo-tions and structural consequences.
3.5 Drops Within Region 5 3.5.1 Plenum Assembly Drop Onto the South Covity Floor The core support assembly is stored in the deeper, south end of the refueling cavity during major repairs; however, on occasion the plenum assembly may also be stored in this location. The potential drop of the core support assembly does 1
not represent a concern because the reactor is defueled prior to the time that a l
drop could occur. A postulated drop of the plenum assembly onto the elevation 550'-6" south cavity floor represents the bounding heavy load.
The south cavity floor is supported by solid reinforced concrete down to the torispherically shaped basemot. A load path for a drop exists directly to the basemat without the potential for local or overall structural response consequen-Therefore, the consequences of a postulated drop of the plenum assembly ces.
onto the south cavity floor are negligible.
3.6 Drops Within Region 6 Heavy loods cannot be carried directly over Region 6 because of physical interferences from the elevation 653' slabs above. The region includes both east and west piping enclosures which are located directly under the elevation 603' slab. Potential impact into the region can only occur os a result of secondary impacts resulting from damage due to perforation or scabbing of the elevation 653' slob.
In addition to protection by the elevation 653' slabs, the piping enclosures are protected by a 2-foot thick slab at their tops. This slab provides significant B-22
protection for concrete that could potentially be ejected from above and miscellaneous items that may be rolled over this area.
Notwithstanding these factors, the systems evaluations conservatively consi-dered a loss of equipment within the enclosures (see Appendix A, Sections 2.l.6 and 2.1.7).
3.7 Drops within Region 7 Heavy loads cannot be carried directy over Region 7 because of physical interference from the elevation 653' slab above. The region includes grating at elevation 603', which provides access to o variety of equipment (see Appendix A, Section 2.1.9). Potential impact into the region can only occur as a result of secondary impacts resulting from damage due to perforation or scabbing of the elevation 653' slob. Toble 5 was evaluated to determine the bounding heavy load carried over elevation 653'in this area.
3.7.1 In-Core Instrument Tonk Access Hatch Covers Drop on Elevation 653' Floor The in-core instrument tank access hatch covers are stored at elevation 653' in on area that directly projects over the Region 7 grating. An evaluation was made to ossess the potential for perforation or scabbing of the 8-inch elevation 653' slab due to a drop of a hatch cover from the normal maximum carry height of 2h feet. The methodology and criterio documented in Appendix B, Section 2.1, Local Impact Response Evaluation, were used in the evaluation.
We have concluded that the slab will not be perforated due to a postulated hatch cover drop; however, scabbing is possible.
Based upon this evoluotion, the systems evaluations (Appendix A, Section 2.1.9) have conservatively considered the loss of equipment due to the impoet of ejected concrete from elevation 653'.
B-23
3.8 Drops Within Region 8 3.8.1 In-Core Instrument Tonk Access Hatch Cover Drop Onto Elevation 606' Floor An evoluotion was conducted to assess the potential consequences of a postu-lated drop of on in-core instrument tank access hatch cover from its position at the hatch penetration at elevation 653' onto the elevation 606' concrete slab which is over the in-core instrument tank.
The methodology and criterio documented in Section 2.1 were utilized to ossess the potential for perforation or scobbing.
The elevation 606' slab is l'-10" thick with a rectangular opening (2'-6" by Il'-6")
that provides access to the in-core instrument tank below.
The opening is covered by a k-inch checkered steel plate.
We have concluded that perforation of the slab will not occur; however, scabbing is possible.
The effects of scobbing on the in-core instrument tank were evoluoted and found to be insignificant.
It is physically possible for a hatch cover to align perfectly during a postulated drop such that it impacts the %-inch steel cover plate and penetrates into the in-core instrument tank. While it is highly improbable that a hatch cover would directly impoet the in-core instrument tank through the elevation 606' penetro-tion, the systems evoluotions (see Appendix A, Section 2.1.10) have conservative-ly considered this impact and have assumed domoge to the contents of the tank.
3.9 Drops Within Region 9 3.9.1 Miscellaneous Drops in the Vicinity of Equipment Hatch items were identified from Table 5 that are corried within Region 9, which is generally the vicinity of the equipment hatch. Under routine circumstances, the automatic reoctor inspection system equipment is the controlling heavy lood handled in this region; however, it is possible during major repairs that a reactor coolant pump motor would be corried out of containment through Region 9.
B-24
While significant protection is provided by the elevation 603' slab to equipment below for postulated drops of most heavy loads handled in this region, our structural experience indicates that it would be resource-effective not to analyze for a drop of a reactor coolant pump.
Accordingly, the systems evoluotions have assumed a loss of equipment below elevation 603' for a projection through this region (Reference Appendix A, Section 2.1.11).
4.0 STRUCTURAL LOAD DROP EVALUATIONS OUTSIDE CONTAINMENT -
RESULTS, CONCLUSIONS, RECOMMENDATIONS 4.1 Component Cooling Water Pump Drop Inside CCW Pump Rom A structural evaluation was completed to ossess the potential for perforation or scabbing of the 2' thick concrete slab which supports the CCW pumps at elevation S78' for o postulated drop of a CCW pump from the highest carry height attainable by the monorail. The methodology and criterio documented in Section 2.1 were utilized. Based upon this evaluation, it was concluded that perforation and scabbing were not probable; however, if scabbing were to occur, the effects were found to be insignificant.
4.2 Service Water Pump Motor Drop in Vicinity of intoke Structure A structural evaluation was completed to assess the potential for perforation or scabbing of the 2' thick concrete roof slab of the service water (SW) valve room area between the intoke structure and pipe tunnel for o postulated drop of a SW pump motor of 7'. This height is consistent with normal handling procedures and corresponds to the maximum height required for the motor to clear its access hatch at the lower intake structure roof and be placed at grade.
The methodology and criterio documented in Section 2.1 were utilized. Based upon this evoluotion, it was concluded that perforation and scobbing were not proboble.
4.3 Spent Fuel Pool Gate Drop Inside Spent Fuel Pool A structural evaluation was completed to ossess the potential for perforation, scabbing, and leakage of the 5' thick spent fuel pool (SFP) base slab due to a postulated drop of a SFP divider gate. The steel divider gates weigh approxi-motely 8,000 pounds and are 25' long,2' wide, and 4" thick. A drop of 42' through air was conservatively assumed, neglecting the energy dissipated by drag in the SFP water. The worst configuration for potential perforation of the pool floor was assumed neglecting the structural resistance of the steel liner plate. The methodology and criterio documented in Section 2.1 were utilized.
It was concluded that perforation and scabbing of the pool floor were not probable.
Furthermore, penetration of the floor slab was predicted to be insignificant.
Based upon this evaluation, leakoge of the pool resulting from a postulated SFP divider gate drop from its maximum carry height is not expected.
B-26
TABLEI POLAR CRAE WAVY LOADSI APPROX.
APPLICABLE WEIGHT OPERATING LIFTING LOAD (POUNDS)
PROCEDURES EQUIPMENT l.
Reactor Plenum Assembly 2 l19,000 SPI 104.46S Plenum Assembly SPIS05.017 Lif ting Rig 2 2.
Reactor Vessel Head 3 330,000 SPI 104.46 Reactor Vessel SPl504.016 Head Lifting Rig 3.
Internals indexing Fixture 4 31,100 SPl 104.46 Plenum Assembly Lifting Rig 2 4.
Plenum Assembly Lif ting Rig 18,S00 SPI 104.46 N/A SPl50S.01 5.
Automatic Reactor 32,000 SPI 104.46 Slings and Shackles Inspection System (ARIS) 6.
I-Beam D-Ring Grating 12,000 spi l04.46 Wire Rope Slings Supports and Shackles 7.
Steam Generator Removable 14,100 SPI 104.46 Wire Rope Slings Supports Supporis and Shackles 8.
Reactor Missile Shields (6) 94,S00 ea.
M-73 Reactor Missile SPI 104.46 Shield Lifting Harness 9.
Polar Crane Load Block i1,200 SPI 104.46 N/A
- 10. Steel Wcrking Platform 2,700 SPI 104.46 Wire Rope Slings and Shackles l1. 20" Steam Generator Snubbers 7,000 SPI 104.46 Wire Rope Slings and Shackles
- 12. Irradiated Specimen Cask 6,000 SPI 104.46 Wire Rope Slings and Shackles 6
TABLEI (continued)
APPROX.
APPLICABLE WElGHT OPERATING LIFTING LOAD (POUNDS)
PROCEDURES EQUIPMENT
- 13. Letdown Coolers 5,000 SPI 104.46 Wire Rope Slings and Shockies 14 Equipment Hatch Covers (Region 2)
SPI 104.46 Wire Rope Slings o 603' el. - 4 covers for I hatch 10,000 eo.
o S85' el. - 2 covers for a hatch 32,000 eo.
- 15. Core Flooding Tank Hatch 8,000 eo.
SPI 104.46 Wire Rope Slings Covers for each of 2 hatches and Shockles (Regions 2 and 9)
- 16. In-Core instrument Tonk 6,800 eo.
SPI 104.46 Wire Rope Slings Access Hatch Covers at and Shockles 603' el. (Region 8)
- 17. Motor Removal Hatches at 5,000 eo. SPI 104.46 Wire Rope Slings 603' el. - l cover for each and Shackles of 2 hatches (Region 2)
- 18. Plenum Assembly Stand 6,000 SPI 104.46 Wire Rope Slings and Shackles
- 19. Core Support Borrel Stand 6,500 SPI 104.46 Wire Rope Slings and Shockles
- 20. Reoctor Vessel Head 12,000 SPI 104.46 N/A Lif ting Rig SPl504.01
- 21. Reactor Coolant Pump (RCP) 4,000 SPl 104.46 Wire Rope Slings Rotating Element and Shockles
- 22. RCP Motor 102,000 spi l04.46 Wire Rope Slings and Shockles
- 23. Reactor Coolont Pump 84,000 SPI 104.46 Wire Rope Slings and Shackles
- 24. Core Support Assembly 224,000 spi l04.46 Core Support SPISOS.01 Assembly
- 25. Reactor Cavity Seal Ring 3,000 spi l04.46 Wire Rope Slings and Shockles
TABLEI (continued)
APPROX.
APPLICABLE WElGHT OPERATING LIFTING LOAD (POUNDS)
PROCEDURES EQUIPMENT
- 26. Refueling Canal Walkways (2) 2,700 SPI 104.46 Wire Rope Slings and Shackles 1.
For reference, o heavy load is defined as the weight of a fuel assembly plus its handling tool; 2,430 lbs.
2.
The total weight of the Plenum Assembly lift includes:
the Plenum Assembly, and the Plenum Assembly Lifting Rig. This lifting rig includes the following components: Head and Internals Handling Fixture; Head and Internals Handling Extension; and the Internals Handling Adapter, Pendants, and Sprecder Ring.
3.
The total weight of the Reactor Vessel head lift includes: the head; the service structure; studs, nuts, and washers; and the Reactor Vessel Head Lifting Rig. This lifting rig includes the following components: Head and internals Handling Fixture, Head and internals Handling Extension, two turnbuckle pendants, and three head lifting cables.
4.
The total weight of the Internals Indexing Fixture includes: the Indexing Fixture and the Plenum Assembly Lifting Rig.
5.
SP l104.46, " Polar Crane System Procedure."
6.
SP 1504.01, " Reactor Vessel Closure Head Removal and Replacement."
7.
SP 1505.01, " Reactor Internals Removal and Replacement."
TABLE 2 COMPOENT COOLING WATER PUMP MONORAll(3) WAVY LOADS APPROX.
APPLICABLE WEIGHT OPERATING LIFTING LOAD (POUNDS)
PROCEDURES EQUIPMENT Component Cooling Water S,400 spi l04.13 1 Slings and Pumps Shackles Component Cooling Water 4,800 SPI 104.13 1 Slings and Pump Motors Shackles 1
1.
SPI 104.13, " Component Cooling Water Pumps Monorail System Procedure,"
includes sufficient discussion to alert maintenance personnel to the safety concern. In addition, it contains sufficient prerequisites and precautions to assure that the operator is qualified, that the monoroll, lifting equipment, and hoist have been inspected and maintained, that the lif ting equipment selected is of sufficient capacity, and that proper system isolation has been effected.
1 I
l TABLE 3 l
SPENT FUEL CASK CRANE HEAVY LOADS APPROX.
APPLICABLE WEIGHT OPERATING LIFTING LOAD (POUNDS)
PROCEDURES EQUIPMENT Pool Divider Gates 8,000 spi l04.502 Wire Rope M-543 Slings and Shackles 1.
This crone is used to handle foods in the Equipment and Fuel Handling Area and for foods around the spent fuel pool. It will eventually be used to handle spent fuel casks. Loads lif ted in the Equipment and Fuel Handling Area are not corried near the spent fuel pool and, therefore, are not listed in this table.
2.
SP 1104.50, " Spent Fuel Cask Crane Operating Procedure."
3.
Maintenance Instruction M-54, Removol/ Replacement of Gate Between ti.c Spent Fuel, New Fuel Storage / Transfer Pools.
e
TABLE 4 INTAKE GANTRY CRAlf FEAVY LOADSI APPROX.
APPLICABLE WEIGHT OPERATING LIFTING LOAD (POUNDS)
PROCEDURES EQUIPMENT l.
Service Water Pump 7,800 SP I 104.531 All intake Gontry Crane heavy foods are lif ted with wire rope slings and shackles 2.
Service Water Pump Motor 8,600 SPI 104.53 3.
Cire Water Makeup Pump 5,700 spi l04.53 4.
Makeup Pump Motor 3,500 SPI 104.53 5.
Roof Top Hatch Covers 2,800 SPI 104.53 6.
Dilution Pump 9,500 SPI 104.53 7.
Dilution Pump Motor 3,000 SPI 104.53 8.
Diesel Fire Pump 5,200 SPI 104.53 9.
Diesel Fire Pump Motor 3,500 SP I 104.53
- 10. Screen Wash Pump 3,560 SPl 104.53 ll. Screen Wash Pump Motor 1,300 spi l04.53 1.
SPI 104.53, "Intoke Contry Crane System Procedure."
TABLE 5 POLAR CRAPE FEAVY LOADS HAPOLED BY REGION INSIDE CONTAINMENT APPROXIMATE REGION (FROM FIGURES I THROUGH 9)
WElGHT LOAD (POUNDS)I 1,*
2 3
4_
S 6
7 8
9_
l.
Reacter Plenum Assembly 2 ll9,000 o#
c c
a a
c c
c c
2.
Reactor Vessel Head 3 330,000 a
a e
a e
c c
c c
3.
Internals Indexing Fixture 4 31,100 o#
c b
o e
e c
c c
4.
Plenum Assembly Lifting Rig 18,500 o#
c o
o a
e e
c c
S.
Automatic Reactor Inspection System 32,000 a#
c c
o o
e c
e a
(ARIS) 6.
I-Beam D-Ring Grating Supports 12,000 e
e b
c c
c c
c c
7.
Steam Generator Removable Supports 14,100 e
e b
c c
c c
c c
8.
Reactor Missile Shields (6) 94,S00/eo.
a+
c a
a c
c c
c c
9.
Polar Crane Load Block 11,200 a+# a a
a o
e e
a a
- 10. Steel Working Platform 2,700 c
c o
o a
e c
e c
II. 20" Ste'am Generator Snubbers 7,000 c
b b
c c
c c
c b
- 12. Irrodiated Specimen Cask 6,000 e
o e
o a
c c
c o
- 13. Letdown Coolers S,000 e
b c
c c
c c
c b
TABLE 5 (continued)
APPROXIMATE REGION (FROM FIGURES I THROUGH 9)
WEIGHT LO_AD (POUNDS) l*
2_
3 4
5_
6 7,
8 9_
c b
c c
c c
c c
c
- 14. Equipment Hatch Covers (Region 2) o at 603' el. - 4 covers for I hatch 10,000/co.
o at 585' el.
- 2 covers for I hatch 32,000/eo.
- 15. Core Flooding Tank Hatch Covers -
8,000/eo.
c b
c c
c c
c c
b 4 covers for each of 2 hatches (Regions 2 and 9)
- 16. In-Core instrumentation Tonk Access 6,800/eo.
c c
a c
c c
o o
e Hatch Covers at 653' el. (Region 8)
- 17. Motor Removal Hatches at 603' el.
5,000/eo.
c b
c c
c c
c c
c l cover for each of 2 hatches (Region 2)
- 18. Plenum Assembly Stand 6,000 e
e c
b o
e c
c c
- 19. Core Support Barrel Stand 6,500 e
e o
b b
c c
c c
- 20. Reactor Vessel Head Lif ting Rig 12,000 o+
o o
o e
c c
e c
- 21. Reactor Coolont Pump (RCP) 4,000 e
b b
c c
c c
c b
Rotating Element
- 22. RCP Motor 102,000 e
b b
b b
c c
c b
- 23. Reactor Coolant Pump 84,000 c
b b
b b
c c
c b
- 24. Core Support Assembly 224,000 o#
c c
b b
c c
c c
- 25. Reactor Cavity Seal Ring 3,000 o#
c c
o e
e c
e c
- 26. Refueling Canal Wolkways (2) 2,700 a+
c o
o c
c c
c c
TABLE 5-
- (continued)
KEY n - lood normally handled in region b - tood potentially handled in region under nonroutine circum. stances (e.g., major repairs) e - lood oc+ expected to be handled in region s
FOOTNOTES U
For {tegion 1: + - RPV head on 7 # - RPV head removed
- l. For reference, the weight of a fuel ossembly plus handling tool is 2,430 lbs.
s
- 2. The total weight of the Plenum Assembly lif t includes: the Plenum Assembly and the Plenum Assembly Lifting Rig. This lifting rig includes the following components: Head and Internois Handling Fixture; Heod and Internals Handling Extension; and the internals Handling Adapter, pendants, and Spreoder Ring.
- 3. The total weight of the Reactor Vessel Head lif t includes: the head; the service structure; studs, nuts and wu:hers; and the Reactor Vessel Head Lifting Rig. This lifting rig includes the following components: Head and Internals Handling Fixture, 4 ead and Internals Handling Extension, two turnbuckle pendants, and three head lif ting cobles.
H
, q 4., The total weight of the Internals indexing Fixture includes: the Indexing Fixture and the Plenum Assembly Lif ting Rig.
5 r
s.
6 h
,_c N
9
TABLE 6
SUMMARY
OF CONTROLLING STRUCTURAL BEHAVIOR RESULTING FROM POSTULATED POLAR CRAE NAVY LOAD DROPS APPROXIMATE WEIGHT OVERALL LOAD (POUNDS)l STRUCTURAL LOCAL 1.
Reactor Plenum Assembly 2 119,000 X
2.
Reactor Vessel Head 3 330,000 X
3.
Internals Indexing Fixture 4 31,100 X
4.
Plenum Assembly Lif ting Rig 18,500 X
X S.
Automatic Reactor Inspection 32,000 X
X System (ARIS) 6.
l-Beam D-Ring Groting Supports 12,000 X
7.
Steam Generator Removable 14,100 X
Supports 8.-
Reqctor Missile Shields (6) 94,S00/eo.
X 9.
Polar Crone Load Block iI,200 X
- 10. Steel Working Plotform 2,700 X
X l1. 20" Steam Generator Snubbers 7,000 X
- 12. Irradiated Specimen Cask 6,000 X
- 13. Letdown Coolers S,000 X
- 14. Equipment Hatch Covers (Region 2) e at 603' el. - 4 covers for I hatch 10,000/co.
X e at S8S' el. - 2 covers for I hatch 32,000/co.
X IS. Core Flooding Tonk Hatch Covers -
X 4 covers for each of 2 hatches (Regions 2 and 9)
- 16. In-Core instrumentation Tonk Access 6,800/co.
X X
Hatch Covers at 603' el. (Region 8)
TABLE 6 (continued)
APPROXIMATE WEIGHT OVERALL LOAD (POUNDS)
STRUCTURAL LOCAL
- 17. Motor Removal Hatches at 603' el. -
5,000/ea.
X l cover for each of 2 hatches (Region 2) i
- 18. Plenum Assembly Stand 6,000 X
- 19. Core Support Barrel Stand 6,500 X
- 20. Reactor Vessel Head Lif ting Rig 12,000 X
X
- 21. Reactor Coolant Pump (RCP) 4,000 X
Rotating Element
- 22. RCP Motor 102,000 X
- 23. Reactor Coolant Pump 84,000 X
- 24. Core Support Assembly 224,000 X
- 25. Reactor Cavity Seal Ring 3,000 X
- 26. Refueling Canal Walkways (2) 2,700 X
FOOTNOTES 1.
For reference, the weight of a fuel assembly plus handling tool is 2,430 lbs.
2.
The total weight of the Plenum Assembly lift includes: the Plenum Assembly and the Plenum Assembly Lifting Rig. This lifting rig includes the following components:
Head and Internals Handling Fixture; Head and Internals Handling Extension; and the Internals Handling Adapter, pendants, and Spreader Ring.
3.
The total weight of the Reactor Vessel Head lift includes: the head; the service structure; studs, nuts and washers; and the Reactor Vessel Head Lifting Rig. This lif ting rig includes the following components: Head and Internals Handling Fixture, Head and Internals Handling Extension, two turnbuckle pendants, and three head lif ting cables.
4.
The total weight of the Internals indexing Fixture includes: the Indexing Fixture and the Plenum Assembly Lif ting Rig.
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FIGURE 3 POSTBUCKLING DEFORMED SHAPE OF FUEL ROD BETWEEN SPACER GRIDS w
,.s
CROSS-SECTION OF STRESS STRAIN FUEL ROD DISTRIBUTION DISTRIBUTION F
Ec y
Compression q9 7
_ R Zone
/
4 Neutral
[
D
-f' Axis J
4 f
y Tension f'
u Zone p
e, FIGURE 4 STRESS-STRAIN DISTRIBUTION TFROUGH CROSS-SECTION OF FUEL LOAD UNDER COMBINED AXIAL COMPRESSION AND BENDING LOADINGS e
Per-Euler Buckling Load tod cto Shaded Area Corresponds to y
Strain Energy
\\
150 5
l 0.05 0.10 0.15 0.18 0.20 AXIAL DEFLECTION a (IN)
FIGURE 5 POSTBUCKLING LOAD-DEFLECTION DIAGRAM FOR FUEL ROD UtOER COMPRESSION
.~z -
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3.
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w 10.
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l 3
e
W 30.
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i 39.
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l
.