RBG-19-354, Forwards Response to NRC 801222 Request for Info in Sections 2.2,2.3 & 2.4 Re Planned Controls for Handling Heavy Loads. SER Outstanding Item 12 Satisfied
| ML20108A673 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 11/05/1984 |
| From: | Booker J GULF STATES UTILITIES CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR RBG-19-354, NUDOCS 8411150119 | |
| Download: ML20108A673 (53) | |
Text
{{#Wiki_filter:p N-A GULF STATES UTILITIES COMPANY P C S T. 0 F F I C E B O X 2 9 51 *DEAUMONT. TEXAS 77704 AREA C O D E 713 838-6631 November 5, 1984 RBG- 19,354
.. File Nos. G9.5, G9.19.2 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation
'U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Denton:
River Bend Station - Unit 1 Docket No. 50-458
-Gulf States Utilities Company (GSU) provides in Attachment 2 the information requested in Sections 2.2, 2.3 and 2.4 of Enclosure 3 to the Nuclear Regulatory. Commissions' (NRC) December ~ 22, 1980 letter. The letter requested a review of the planned controls for handling heavy
-loads at River Bend Station (RBS). GSU has performed a comparison of these controls to the guidelines of NUREG-0612 to demonstrate that the guidelines have been or will be met and to satisfy the NRC RBS Safety
. Evaluation Report outstanding item No. (12).
~0n March 1, 1984 GSU submitted.a report addressing the first phase of the requested review (Section 2.1 of Enclosure 3 to the NRC's December 22, 1980 letter). A technical evaluation report.was performed by EG&G Idaho, Inc. on GSU's phase one submittal. As a result of their
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review, additional information was requested in a NRC letter dated September 14, 1984. GSU's.responr.e to the NRC's request for additional information is addressed in Attachment 1. Sincerely,
- g. 2. W J. E. Booker Manager-Engineering, Nuclear Fuels & Licensing River Bend Nuclear Group Attachment 8411150119 841105 gDRADOCK 05000458 PDR I i
l i w _
y . - , ATTACHMENT 1 j- Guideline 1, NUREG-0612, Article 5.1.1(11: NRC Request: Information is needed to confirm that for other heavy loads handled by the Polar crane and the other nonexempt cranes or hoists (35 identified loads), safe load paths are followed. Additional evaluation and information, as discussed, is needed for RBS to show consistency.with NUREG-0612 Guideline 1, for 35 loads. GSU Response: Of 'the 35 identified loads, Safe Load Path u designations / restrictions, were necessary for approximately 17 loads handled by three cranes and two monorails, and are presently being incorporated into RBS load handling procedures. The approach for this selection is based on the conclusions found in the phase II report (note pgs. 1-5, 29-31). Guideline 2. NUREG-0612, Article 5.1.1(2): NRC Request: The information submitted provides a commitment to develop procedures. However, the five specific requirements that these procedures should include (see 2.3.2 above) have not been addressed. In the preparation RBS should include all of the requirements or justify exceptions. Suitable resolution of the Safe Load Path guidelines must be established before procedures consistent with Guidelines can be written. At a minimum, procedures should cover handling of those loads listed in Table 3-1 of NUREG-0612. These procedures should include: identification of required equipment; inspections and acceptance criteria required before movement of load; the steps and proper sequence to be followed in handling the load; defining the safe path; and other special precautions. c RBS should supplement the commitment to develop procedures with information or statements to confirm that the procedures incorporate all of the requirements specified in NUREG-0612 Section 5.1.1(2). GSU Response: GSU is developing Load Handling Procedures for systems of concern to ccr.tain the requirements of 5.1.1(2).
Guideline 4, NUREG-0612, Article 5.1.1(4) a NRC Request: 1 Confirm whether there are three or only two "Special Lifting Devices." If in fact there are three identify the third one and provide information to verify its consistency with NUREG-0612 Guideline 4 requirements. GSU Responce: Only two special lifting devices exist as described in phase I. The third lifting device was removed from the RBS design during phase I evaluation. Based on phase II evaluation results, the five loads handled by the special lifting. devices are not " critical loads" as defined by ANSI N14.6. Guideline 7, NUREG-0612, Article 5.1.1(7) l NRC Request: A suitable resolution of the inconsistent exception concerning initial testing will bring RBS into acceptable consistency with the NUREG-0612 Guideline 7. ' s GSU Response: There is no longer any exception to initial crane testing identified since test loaded movement of bridge and trolley are included in RBS start-up procedures for the three cranes of concern. V -Q- _u
0 1
- p. RESPONSES TO REQUESTS FOR IWORMATION
( ) IN SECTIONS 2.2,2.3, Abo 2.4 OF ENCLOSURE 3 sJ TO NRC DECEMBER 22,1980 LETTER i l 2.2 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN THE VICINITY OF FUEL STORAGE POOL i NUREG-0612, Section 5.l.2, provides guidelines concerning the design and opera-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 area, 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 evaluation criteria of NUREG-0612, Section 5.1, Criteria i through 111. ITEM 2.2.1 Identify by nome, type, capacity, and equipment designator, any cranes physically capable (i.e., ignoring interlocks, moveable mechanical stops, or operating procedures) of carrying loads which could, if dropped, land or fall into the spent fuel pool. (N RESPONSE: At River Bend, there are two spent fuel storage pools. One in the k Fuel Building at the i13' el. capable of storing up to 525% of a full core and another in the Reactor Building capable of storing up to 32% of a full core. (See Figures 7 and 6, Regions 19 and 2 respectively). The spent fuel storage pools and rocks are described in FSAR Section 9.l. With regard to the Fuel Building, there are no cranes capable of carrying heavy loads over the spent fuel pool. In the Reactor Building, the Polar Crane is capable of carrying loads over the spent fuel storage area. It is on overhead bridge crane mounted on a circular rail with a main and auxiliary hoist with capacities of 100 and S tons, respectively. l l l l l ' p]s L l DC-84-13 l l
ITEM 2.2.2 Justify the exclusion of any cranes in this area from the p) x" above category by verifying that they are incapcble of carrying heavy loods or are permanently prevented from movement of the hook center!!ne closer than 15 feet to the pool boundary, or by providing a suitable analysis demonstrating that for any failure mode, no heavy load con fall into the fuel-storage pool. RESPONSE: With regard to the Fuel Building spent fuel storage creo, the Spent Fuel Cask Trolley is located at el 151'. However, the spent fuel pool is protected from drops from this crane by restriction (i.e., location of rocks) of the limits of
- cask crane travel as discussed in subsections 9.l.4.2.2.1 of the River Bend FSAR.
The Fuel Building Bridge Crane will be provided with fixed mechanical stops prior to storage of any spent fuel in the spent fuel pool. These stops will restrict crane movement such ihot it is incapable of carrying heavy loads over spent fuel. With regard to both the Fuel Building pool and the Reactor Building spent fuel storage creo, there are 1/2 ton capacity jib cranes that can lift loads over the spent fuel storage rocks. As indicated in our initial response, however, these cranes are only used to carry lighter loads, such as channels, control rods or fuel assemblies that do not qualify as heavy loads. b! v DC-84-13 2
ITEM 2.2.3 Identify any cranes listed in 2.2.I, 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.l.6 or partial compliance supplemented by suitable , alternative or additional design features). For each crane so evaluated, . provide the load-handling-system (i.e., crane-lood-combination) information specified in Attachment 1.
RESPONSE:~ lt has not been necessary to evaluate the Polar Crane against the criteria of NUREG-0612, Section 5.l.6. DC-84-13 3
I ITEM 2.2.4 For cranes identified in 2.2.1, above, not categorized l fw occording to 2.2-3, demonstrate that the criteria of NUREG-0612, l ( ) Section 5.I, are satisfied. Compliance with Criterion IV will be V demonstrated in response to Section 2.4 of this request. With respect to Criterio I through lli, provide o discussion of your evaluation of crane operation in the spent fuel area and your determination of compliance. This response should include the following information for each crane:
- a. Which alternatives (e.g.,2,3, or 4) from those identified 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 mechanical stops and indicate the circumstances, if any, under which these protective devices may be bypassed or removed. Discuss any administrative pro m-dures invoked to ensure proper authorization of bypass or removal, and provide any related or proposed technical specification (operation-of 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 (p)
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storage of certain quantities of spent fuel at specific post-irradiation decoy times, provide present and/or proposed technical specifica-tions ad discuss administrative or physical controls provided to ensure that these assump-tions remain valid.
- d. Where reliance is placed on the physical loca-tion of specific fuel modules at certain post-irrodiotion decay times, provide present and/
or proposed technical specifications and discuss administrative or physical controls provided to ensure that these assumptions remoir, valid.
- e. Analyses performed to demonstrate compli-ance with Criterio I through lit should conform to the guidelines of NUREG-0612, Appendix A.
Justify any exception taken to these guide-lines, and provide the specific information requested in Attachment 2, 3, or 4, as appro-priate, for each analysis performed. C 1 DC-84-13 4
l 4 RESPONSE: The heavy loods that could be handled by the Polar Crane were identified in Table 4 in GSU's initial submittal to NRC regarding the heavy loads
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issue. As indicated in the response to item 3.o of that submittal, both procedural restrictions and Technical Specifications have been developed to prevent carrying heavy foods over spent fuel in the rocks of the Containment fuel pool. Tnerefore the requirements of Criterio I through fli of NUREG-0612 are satisfied. I O i tO
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DC-84-13 5 l
. 2.3 SPECIFIC REQUIREMENTS OF OVERHEAD HANDLING SYSTEMS '
OPERATING IN THE CONT AINMENT {'} y NUREG-0612, Section 5.l.3, provides guidelines concerning the design and opero-tion of load-handling systems in the vicinity of the reactor core. Information provided in response to this section.should be sufficient to demonstrate that
. odequate measures have been token to ensure that in this area, 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 evaluation criteria of NUREG-0612, Section 5.1, Criteria i through 111.
11EM 2.3.1 Identify by name, type, capacity, and equipment designator any cranes. physically capable (i.e., taking no credit for any interlocks or operating procedures) of carrying heavy loads over . .the reactor vessel. 4 RESPONSE: The only handling system within containment physically capable of carrying heavy loads over the reactor vessel is the Containment Polar Crane. A 1.] t O DC-84-13 6
l l ITEM 2.3.2 Justify the exclusion of any crones in this area from the ('N 3 above category by verifying that they are incapable of carrying heavy loads, or. ore permanently prevented from the movement of any load
'V-either directly over the reactor vessel or to such a location where in the event of any lood-handling system failure, the load may land in or on the reactor vessel.
RESPONSE: The only other handling system inside the containment capable of moving loads over the vessel is the Refueling Platform used for refueling operations. Its' function is to handle single fuel assemblies and perform other vessel servicing functions, i.e., no heavy foods os defined in NUREG-0612 ore handled by this handling system. l N d l~ O DC-84-13 7
ITEM 2.3.3 Identify any cranes listed in 2.3.1 above which you have [N, evoluoted as having sufficient design features to make the likelihood \, ! of a load drop extremely small for all loads to be corried and the basis for this evoluotion (i.e., complete compliance with NUREG-0612, Section 5.l.6, or partial compliance supplemented by suitable alternative or additional design features). For each crane so evolu-ated, provide the lood-handling-system (i.e., crone-lood-combination) information specified in Attachment 1. RESPONSE: As indico. :d in the response to item 2.2.3 above, it has not been found to be necessary to evoluote the Polar Crane against the criteria of NUREG-0612, Section 5.1.6.
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DC-84-13 8
_ ITEM 2.3.4 For cranes identified in 2.3.1 above not categorized TN according to 2.3.3, demonstrate that the evaluation criteria of \j. 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 crane operation in the containment and your determination of compliance. This response should include the following information for each crane: ITEM 2.3.4.a. Where reliance is placed on the installation and use of electrical interiocks 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 crane load blocks have 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 heavy load. The load block is used for handling numerous loads, including the reactor vessel (n) head, drywell head, steam dryer, and moisture separator. in moving these loads, the hook, load block, rope, drum, sheave assembly, motor shafts, gears, and other load bearing members are subjected to significant stresses approaching the load rating of the crane. By comparison, these components are subjected to a considerably smaller load when only the hook and load block are being moved. Based on this, it is not considered feasible to postulate a random mechanical failure of the crane load bearing component: when moving either the main hoist or auxiliary hoist load block without a food. 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 a "two blocking" position with continued hoisting by the motor and subsequent parting of the rope (this situation can be prevented by operator action prior to "two blocking" or by on upper limit switch to terminate hoisting prior to "two blocking"); and rO DC-84-13 9
- 2) Uncontrolled lowering of the load block due to failure of T' the holding broke to function (the likelihood of this con be (N / made small by use of redundant holding brakes).
The River Bend polar crane main hoist and auxiliary hoist are each provided with redundant and diverse upper limit switches to interrupt power to the hoist motor prior to "two blocking". When power is removed, holding brakes are automatically applied. The holding brakes are solenoid released, and spring opplied on loss of power to the solenoid. Two holding brakes are provided for each hoist on the polar crane; each holding broke hos sufficient capacity to hold the rated load. Additionally, inspection procedures assure that the limit switches and holding brakes are functional and properly adjusted. With the provisions described above, the redundant 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 load block and hook is of sufficiently low (] - likelihood that it does not require load drop analyses. 4 (~\ l G L i l DC-84-13 10
ITEM ~ 2.3.4.b. Where reliance is placed on other, site-specific
/N considerations (e.g., refueling sequencing), provide present or
() proposed technical specifications and discuss administrative or physical . controls provided to ensure the continued validitiy of such considerations. RESPONSE: Loads only lifted over the vessel when the reactor vessel head or moisture separator is in n! ace were not considered as loads that could potentially drop into the core. These are: the drywell head and the steam dryer. No administrative controls are required to enforce this situation, because it is physically impossible to disassemble or reassemble the reactor such that these loads would be corried over an exposed core. In addition, the portable refueling shield is installed in the reactor well after the head has been removed, but before the dryer or separator has been removed. It is removed from the reactor well after the dryer and separator have been installed. This sequencing is enforced by written procedures governing the installation and removal of the portable refueling shield and will be strictly enforced by individuals in charge of lifts by the Polar Crone. (3 - h n , V DC-84-13 lI
ITEM 2.3.4.c Analyses performed to demonstrate compliance with
-m Criteria 1 through 111 should conform to the guidelines of NUREG-0612, Appendix A. Justify any exception taken to these guidelines, (d) and provide the specific information requested in Attochment 2,3, or 4, as appropriate, for each analysis performed.
RESPONSE: There are three potential consequences of interest when consider-ing load drops onto the open reactor vessel. They are: 1) loss of reactor vessel integrity,2) fuel clodding damage and the resultant radiological dose, and 3) fuel crushing and the possibility of a resulting criticality condition. Criteria i through "! in Section 5.1 of NUREG-0612 oddress each of these potential consequences. The evaluations below have been performed to address these issues. Analyses were performed to determine the structural consequences of dropping the vessel head or the shroud head assembly during maintenance operations. The consequences of dropping the steam dryer assembly con be extrapolated from the analysis of the shroud head assembly drop, since a steam dryer drop would generate less kinetic energy than the shroud head assembly drop and the impacted structure would be the some for both cases. The shroud head and dryer were assumed to be dropped from a height sufficient to generate the steady state velocity of the two assemblies os they move through water inside the vessel and are under the action of the fluid drog forces. An oxisymmetric impact of the shroud head and dryer assembly on the main body of the shroud is assumed. In addition, nonoxisymmetric impact of the shroud head on the shroud f was also considered. It was postulated that the vessel head would be dropped from a height of approximately 39 ft. obove the vessel-head flange, and that at impact the head would be rotated 900 from the inplace orientation cousing a point impact on the vessel. This height conservatively bounds the maximum possible carry height as limited by physical restrictions. The vessel loods due to the postulated impacts were determined by dynamic, elastic-perfectly plastic finite element analyses. In the vessel head impact
- analysis the vessel was characterized by isoporometric quadrilateral elements using the ANSYS computer program.
. m DC-84-13 12
-- - - - - , -w +w+ - -- - - - ,ym.. -, --ey- -
The model reflects the longitudinal, lateral, extensional and inextensional l ['T effects, in the shroud head impact analysis the impacted shroud body is V characterized by oxisymmetric finite elements using the ANSYS computer program. The.results of the finite element analyses can be suinmarized as follows: Drop of: Consequences Vessel Head Local yielding of the vessel top flange Reactor skirt does not yield, does not buckle, and will remain stable. Vessel maintains its normal position. No damage to the fuel rods and hence no release of radioactive materials. Shroud Head / Steam No yielding of the upper shroud and shroud support Dryer struts. No instability of the shroud support structure. (J (
)
Damage to the internal components is minor and does not impact structural stability. No damage to the fuel rods and hence no release of radioactive materials. In addition drops of the drywell head onto the vessel head and the portable refueling shield onto the separator (dryer conservatively not relied on to mitigate the drop) were evaluated by comparing the available drop energies to those in the analyses described above. Based on this comparison, these drops l were found to be bounded by the head drop and dryer drop analyses. - On the basis of the analyses described above, it is concluded that NUREG-0612 Criteria I-ill are met for all postulated drops into the reactor well. l l l t DC-84-13 13
2.4 ' SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS _ OPERATING IN PLANT' AREAS CONTAINING EQUIPMENT Y (d REQUIRED FOR REACTOR SHUTDOWN, CORE DECAY HEAT REMOVAL, OR SPENT FUEL POOL COOLING NUREG-0612, Section 5.l.5, provides guidelines concerning the design and opero-tion of food-handling systems in the vicinity of equipment or components required for safe reactor shutdown and decay heat removal. Information provided in response to this section should be sufficient to demonstrate that adequate measures have been taken to ensure that in these areas, either the likelihood of a load drop which might prevent safe reactor shutdown or prohibit continued decay heat removal is extremely small or that damage to such equipment from load 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-! and their loads in your response to 2.l.3.3. ITEM 2.4.1 Identify any cranes listed in 2.l.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.l.6, or partial compliance supplemented by suitable alterantive or additional design features). For each crane so evolu-ated, provide the food-handling-system (i.e., crane-load-combination) information specified in Attachment 1.
/3 RESPONSE: The handling systems of interest specified in response to item 2.l.1 are also listed below. It has not been necessary to evaluate any of these handling systems aginst the criteria of NUREG-0612, Section 5.l.6.
Capacity Handling System (Tons) Location Reactor Building Polar Crane / Aux Hoist 100/5 Reactor Building Drywell MSIV and Relief Valve Monorail 3 Reactor Building Fuel Building Bridge Crane 15 Fuel Building I Spent Fuel Cask Trolley / Aux Hoist 125/15 Fuel Building i MSIV Monorails 8/5 Auxiliary Building l MSIV and Feedwater Isolation 3 Auxiliary Building Valve Monorails j Feedwater Valve Hoists 3 Auxiliary Building O V l ! DC-84-13 14 i
Capacity
" """~ "" '""""
i CJ RHR A Pump Monorail 8 Auxiliary ouilding RHR B & C Pump Monorail 8 Auxiliary Building Auxiliary Building Tunnel Plug 6 Auxiliary Building Monorail Hoist Area Monorail 5 Control Building Floor Plug Monorail 5 Control Building Control Building Equipment 5 Control Building Handling Area Monorail 4 O J t f ? I O DC-84-13 15
ITEM 2.4.2 For any cranes identified in 2.1-1 not designated as single-failure-
,, proof in 2.4-1, o comprehensive hazard evaluation should be provided which includes the following information:
(v)
- o. The presentation in a matrix format of all heavy loads and potential impact areas where damage might occur to safety-related equipment. Heavy loads identification should inclu e Asignation and weight or cross-reference to informatica p. wided in 2.1-3-c. Impact areas should be identified by conste etion zones and elevations or by some other method such that the impcct area con be located on the plant general arrangement drawings. Figure I pro-vides a typical matrix.
- b. 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:
(l) For load / target 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-
,q related function). \v )
(2) Where mechanical stops or electrical interlocks are to be provided, present details showing the creas where crone 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 crane use, and for verifying that interlocks are restored to operability offer operations which require bypassing have been completed. (3) Where load /torget combinations are eliminated on the basis of other, site-specific considerations (e.g., maintenance sequencing), provide present and/or proposed technico! specifications and discuss admin-istrative procedures or physical constraints invoked to ensure the continued validity of such considero-tions.
- c. For interactions not eliminated by the analysis 2.4-2-b, above, identify any handling systems for specific loads which you have evcluated as having sufficient design features to make the likelihood of a load drop extremely small and the basis for this evaluation (i.e., complete 5
A i DC-84-13 16
, , . , . . , - - - - - - . - , , . _ , -- . --.. ---- , , _ _ . - . . - - . - - . - a , , . ,
l l compliance with NUREG-0612,. Section - 5.l.6, or partial r compliance supplemented by suitable alternative or addi-('i) tional design features). For each crane so evaluated, provide the lood-handling-system '(i.e., crone-lood-combi-I _( nation) information specified in Attachment I.
- d. For interactions not eliminated in 2.4-2-b or 2.4-2-c, above, demonstrate using appropriate analysis that
~domage would not preclude operation of sufficient equip-ment to allow the system to perform its safety function following a load drop (NUREG-0612, Section 5.1, Criterion IV). For each analysis so conducted, the following infor-motion should be provided.
(1) An indication of whether or not, for the specific load being investigated, the overhead crane-handling system is designed and constructed such that the hoisting system will retain its load in the event of seismic accelerations equivalent to those of a safe shutdown earthquake (SSE). (2) The basis for any exceptions taken to the analytical guidelines of NUREG-0612, Appendix A. (3) The information requested in Attachment 4.
\/v ], RESPONSE: Table I identifies the potential impact regions cod the associated load handling systems. The loads corresponding to these handling systems were previously identified in response to item 2.1-3c. - Figures I through lI further describe the potential impact regions. For the handling systems identified in Table I (also listed in 2.4.1 above) a combination of systems and structural evaluations was utilized to determine if Criterio IV of NUREG-0612 is met for all postulated lood drop scenarios. To assist these evaluations, o set of safety functions were identified corresponding to these criteria. The goal of these evoluotions then become to demonstrate that the applicable safety functions could be accomplished for all load drop scenarios.
SYSTEMS EVALUATION METHODOLOGY As part of the evoluotion of heavy lood handling operations at River Bend Station, o number of potential load drop regions in the Reactor, Auxiliary, Fuel and Control Buildings, were cddressed by performing a " systems evoluotion". O lJ DC-84-13 17
l The objectives of the " systems evoluotions" were to demonstrate that safe shutdown, long-term cooling, and fuel pool cooling cou!d be achieved and/or
/ 'T .
t V) maintained ossuming that certain equipment was lost as a result of postulated l I load drops. In ' order to demonstrate the ability to occomplish these objectives, it was necessary to:
- 1. Identify the load drop regions.
- 2. Identify the safety functions required to be accomplished for each region.
- 3. Identify the systems needed to accomplish the identified safety functions.
- 4. Identify the equipment associated with these systems, including support systems, that could potentially be lost if a load drop were to occur in the region.
- 5. Determine the effects of loss of this equipment on the ability to accomplish the identified safety functions.
p t SAFETY FUNCTIONS The systems evoluotions address the capability to perform the following safety functions:
- 1. Spent Fuel Cooling - upper Reactor Building fuel pool and Fuel Building spent fuel pool.
- 2. Extended Core Cooling.
- 3. Reactor Shutdown and Cooldown to extended . core cooling.
To determine which safety functions are opplicable to each load drop region, initial plant conditions were assumed during load handling operations as discussed below.
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t i DC-84-13 18 l l
Spent Fuel Cooling Load handling operations in any region could take place when spent fuel is stored in the Fuel Building spent fuel pool and the Reactor Building fuel pool. Therefore the capability to perform spent fuel cooling in these storage locations must be verified for postulated load drops in all regions subject to systems evoluotions. The reactor building fuel pool contains spent fuel racks capable of storing up to 32% of a full core. The Fuel Building spent fuel pool has a capacity of 525% of full core. For the purpose of evaluating whether or not this safety function could be accomplished, heavy load drops were postulated to occur when there was recently discharged irradiated fuel in the containment spent fuel storage rocks and in the Fuel Building spent fuel storage rocks. Extended Core Cooling Load handling operatic,ns in any region could occur with fuel in the reactor vessel. Therefore, for all postulated load drops, the capability for extended core cooling must be verified, f3 O Reactor Shutdown and Cooldown to Extended Core Cooling it is not anticipated that any of the heavy loads handled by the Containment Polar Crane or the Drywell MSiV and Relief Valve Monorail would be lifted until the plant has been shut down for some time. Accordir. gly, all heavy load drops inside the Reactor Building were postulated to occur when the reactor was shut down and cooled down. This condition corresponds to either the Cold Shutdown or Refueling Operational Conditions. Load handling operations in the Auxiliary, Fuel, and Control Buildings could take place during any plant operational condition. Accordingly, the capability to perform this safety function must be verified for postulated load drops in regions 10 through 18,20 and 21. SYSTEM REQUIREMENTS For the purpose of performing the systems evaluations described in this appendix, only " safety systems" were selected to accomplish each of the safety , 'V I
- DC-84-13 19 l
r. functions of interest. - This means that many of the systems used to accomplish M normal plant shutdown, such as offsite power nnd condensate, feedwater and b- circulating water systems, were conservatively not relied on to perform the systems evaluations. Further, although there are several combinations of safety systems that may be used for safe shutdown, Initially only certain systems were assumed to be available. Additional systems, or portions of systems, were included in the review only if it was determined to be necessary or prudent to do so. Cooling of spent fuel in the Fuel Building and/or the upper containment fuel pool during maximum postulated heat loads con be accomplished by the Fuel Pool Cooling and Cleanup (FPCCU) system. The fuel pool cooling subsystem of the
.FPCCU consists of two 100-percent capacity pumps and coolers, and associated piping, valves, and instrumentation. Normal cooling of the FPCCU system
~
, coolers is provided by Reactor Plant Component Cooling Water (RPCCW). Upon i foss of RPCCW pressure, and as assumed in these systems evaluations, the . . Standby Service Water (SSW) system provides cooling to the FPCCU heat g exchangers. SSW uses the RPCCW system piping which is isolated from the Q balance of the RPCCW system by redundant (Division I and 2) isolation volves. Although one loop of the FPCCU system con provide odequate spent fuel cooling during maximum design basis heat loads, both FPCCU loops are normally used during reloods. When fuel is in the containment and Fuel Building pools (e.g. reload) one loop of the FPCCU systems is oligned to the Fuel Building and the other loop to containment. If either FPCCU loop fails the operable FPCCU loop is aligned to the Fuel Building and the standby RHR system loop is aligned to containment. j- Figure 12 displays the systems, selected for evaluation purposes, to perform the Reactor Shutdown, Cooldown, and Extended Core Cooling functions. Evaluation of the systems included all necessary support systems. Performance of the safety functions by other safety related systems was considered when required by the potential effects of the postulated load drops. t s I DC-84-13 20 t
, , , , -- ,,. . ..---..---.-,w,-,--.,
LOAD DROP MATRICES
-( )
\d For each safety function, a matrix was developed that outlined each of the regions to be evaluated and the critical system components relied on to perform the safety function. The critical components include required support functions such as cooling water, power supplies, and electrical cabling.
Each . region of concern was evaivated to determine which of the system components of interest could be lost as a result of a load drop. The evoluotion was performed by conducting plant walkdowns and reviewing plant drawings, system descriptions and fire hazards evoluotions. This information was then entered into the matrices. STEPS IN THE SYSTEMS APPROACH The following summarizes the steps that were performed in the systems evaluations for each safety function evaluated:
/7 Completion of Matrix (i) ' Identify the system (including any support systems) com-ponents selected for accomplishing the safety function of
, interest, organize into component groups for purposes of evaluation, and enter at top of matrix columns. l (2) For each region of concern, evaluate the potential for damage / loss of system components based on a detailed-review of the equipment and piping layouts, electric cable j locations and results of fire hazards analysis electrical ! review of the region. Enter the components assumed lost in the appropriate box on the matrix. (3) Compare system equipment required (item I), with equip-ment lost (item 2), and determine if the safety function i for which the system is relied on could be lost. (4) Review for other potential system interactions based on ! equipment damaged / lost and determine if safety function ! could be lost. l l b. IU l DC-84-13 21 ~ -.
l 1 l l l Conclusions V (5) If the system evaluation reveals that the system could accomplish its safety function following a load drop into the region of interest, then no further evoluotion is necessary. (6) If the system evaluation reveals that the system function could potentially be lost, then evaluate the possibility of relying on alternative safety systems to accomplish the same function following a postulated load drop into the region. (7) The overall safety function conclusion regarding a part-icular region is the composite for that region of the conclusions for all the systems required to accomplish the safety function. STRUCTURAL EVALUATION METHODOLOGY Structural load drop analyses performed to support evaluations related to safety
- functions described above, typically involved determination of structural response of concrete floor slabs to dynamic impact loadings. The heavy loads which could potentially be dropped onto various floor sicbs were evoluoted to c. >N identify foods which control local response (e.g. penetration, scabbing, spalling, Q perforation, etc.); loads that control overall structural response (e.g., large inelastic deformations or abrupt failures of principal structural members, etc.);
and/or lor.ds that may induce behavior that exhibits combined response such that either overall or local failure modes would control. Where the controlling mode of response to postulated load drops is listed as
" local", then loads were evaluated to determine the potential for slab penetro-tion or perforation. Scobbing of the concrete deck backface was evaluated for all loads. Where it was found that postulated drops are capable of prkucing this scabbing effect, it was decided that the consequences of scabbing would be considered in the systems evaluations, i.e., its potential for damaging equipment below the floor was evaluated.
Where the controlling mode of response is listed as "overall structural", these food drops were evaluated to determine the potential for producing gross and intolerable distortions of primary structural members and possibly propagating failures. (V i DC-84-13 22
Overall Structural Response Evoluotions I L V A model of each fl 1 tion was developed with the objective of evoluoting s structural behavio. ,astulated flat and oblique drops of these loads. A food drop methodology was developed to investigate the important modes of structural behavior. The objective of this methodology is to characterize strt.ctural 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 characterized 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 can be absorbed by the floor system without reaching those limits of unacceptable behavior is balanced against the externally applied energy resulting from o heavy load drop. It has been assumed that momentum is conserved and the kinetic energy of the drop
- d. ives the mass of the floor and induces strain. As an additional conservatism, no credit was taken for potential sources of energy dissipation such as concrete crushing and penetration.
[)\ L. . l i l l l I l l' DC-84-l 3 23 l _.__i,____. _ . . _ . _ _ _ _ _ _ , _ _ _ _ _ _ _ _ _ . . _ _ _ . _ , _ . _ _ - .
An iterative step-wise linear static analysis was performed for the postulated load d' rops whose controlling mode of response was determined to be "overall ["') V structural." .The objective was to determine force-deflection for important points in the structural model. The computation procedure of the analysis is based on a network interpretation of the governing' equations, the principal _ feature of which is the segmentation in processing of the geometrical, mechani-col and topological relationships of the structure. This allows a concise and systematic computation algorithm that is applicable for different structural types. For each impacted structural system (floor slab or slob-beam composite), a model was developed and the response of the system to the dynamic impact loading was determined. The model was loaded in the direct vicinity of the drop location. This is considered to be conservative in view of the fact that the sicb will help transfer load away from the drop vicinity and result in a more favorable redistribution of the food. The stiffness properties of the supporting beam grid, where applicable, were p represented assuming an effective reinforcement for the beam / slab composite V section consistent with ACI 318-77 (Reference 1). The model was loaded until the moment capacity of any section or the allowable deflection was reached. This moment capacity is defined by Chapter 10 of ACI 318-77 (Reference 1). Generally, the ultimate load of a slob / grid system is reached prior to exceeding , the hinge rotational capacity of particular sections provided that on unstable mechanism has not formed. This was found to be the cose in this analysis. The hinge rotationo! copocity was used as a criterio to set the maximum allowable level of deflection for the slob / grid system. The hinge rotational capacity for concrete structures was developed in References 2 and 3 based on test results given in References 4 and 5 and is given as: A DC-84-13 24
i: r u= 0.0065 (d/c) 5 0.07 (1) (s.^ ) where, ru = rotational capacity of plastic hinge (radions) d = distance from the compression face to the tensile reinforcement, c = distance from the compression face to the neutral axis'of ultimate strength. The maximum deflection for a beam with a plastic hinge at its center, is then given by: Xm = (ru l)/4 (2) where, Xm = maximum deflection, 4,nI L = span of beam .O Rotations of the magnitude governed by equation (1) 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 scobbing 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 l plug formation. The potential for scabbing was evaluated for all load drops. t i l The load / deflection history up to the point of the ultimate loading, coupled with l the maximum allowcble deflection, defines the maximum level of strain energy l l- obsorption, provided that a shear failure has not occurred. The shear stress at I- limiting sections was checked and compared to allowables os specified in Chapter ll of ACI 318-77 (Reference 1). [. For each area where the potential for overall structural response modes was I- considered possible, on ossessment of the bounding drop was made. The criteria
- for selection was impact energy of the postulated drop.
I b DC-84-13 25
.. ~. . . _ _ _ _ _ ._ . _ _ . . . _ .
in addition to the conservatisms previously mentioned, the following conserva-L tisms are also inherent in the methodology used in the evoluotion: Static material strengths for concrete and steel were I) U' used. Test dato shows that this property increases with [, the increased strotri rates ossociated with dynamic lood-W? 'ings. For example, References 6 and 7 recommend dynamic increase factors of 1.25 for the compressive strength of concrete and 1.20 for the flexural, tensile and compressive strength of structural steel. L 2) Design (minimum) material properties for concrete and
- steel were used. No increase was taken for the aging of concrete .which con amount to o factor of up to 1.35 (Reference 8) of increased strength. Also, the overage
, strength- for structural steel is nearly a factor of 1.25 (Reference 9) higher than the minimum yield requirement specified by ASTM. While these factors above minimum code strength exist and contribute to structural margins, they were not used in the evaluation.
- 3) Equation (l)' for hinge rotational capacity was used. This corresponds to support rotations of the order of 2 degrees
, with minimum cracking and no crushing or scabbing. To
, meet necessary performance requirements (i.e. halting propgoting failures), larger rotations in the range of 5 to O
V-- - 12 degrees could be tolerated. Such rotations would lead to crushing, spalling and s-abbing of the section (Reference 7); however overall food carying capability _ is expected to remain intact. Experimental observations
; suggest even further capability for well designed and well
-anchored slobs. Failure modes at such levels initially appear to be controlled by yielding in shear and flexture followed by membrane stretching until ' failure occurs, normally at the support edge of the slab. - Use of these larger rotational capabilities would have resulted in j greater energy absorbing capabilities of the grid system.
- 4) The analysis used ACI 318-77 allowable shear stresses. A
.significont body of dato suggests the existence of higher shear capabilities on the order of 10 Vf'c to 20 Vf'c
~ (References 11-19). It is expected that the shear capabil-
, ities for these beams would tend to be in the higher end of
! the range since the majority of the beams are " deep". !. Deep beams behave os tied-orches with significant re-serve capacity. -
- 5) in many cases, the analysis neglected the two-way resis-tonce capability of the slob. It is expected that the slab
- would contribute increased strength particularly at larger +
- j. deformations.
.k s
DC-84-13 26
- 6) The load was distributed directly under the dropped shield r~n plug. In reality a more favorable load distribution would (f exist due to the load distribution capability of the slab.
No credit was taken for local energy dissipation associ-7) ated with any crushing of the shield plug or the immediate surface of the floor. Local Structural Response Evaluations Selected loafis such as the portable radiation shield (cattle chute), various hatch covers and equipment were evaluated to assess the acceptability and potential consequences of postulated drops. The acceptance criteria were based on the capability of the concrete slabs to resist perforation, penetration, and underside scabbing. Procedures recommended in References 20 and 21 were followed. The modified National Defense Research Committee (NDRC) formula (Reference 22) was chosen because it has been shown to give the best fit with available experimental data (References 23 and 24). The NDRC formula for the depth of penetration, x 3 (inches), of a solid cylindrical missile is given by: (Y x=( 4 KNWd (V)l.8 /(1000 d) )b for x/d 6 2.0 (2) or x = (KNW (V)l.8 / (1000 d) + d for x/dE2.0 (3) 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 p
DC-84-13 27
,, e - - , - - - , ,. ,..,..,rw.,.----n,- -----------,-----,.,r, . - , , - - -,
= 1.00 spherical-nosed missiles
= 1.I4 sharp-nosed missiles O- K = concrete penetrability factor
= 180/d f'c (f'c = concrete compressive strength pounds / square inch The thickness of reinforced concrete needed to recist impact without perforation and scabbing are given by the following Army Corps of Engineers formulae which can be used in conjunction with equations (2) and (3) (Reference 25).
ts/d = 2.12 + l.36 (x/d) for 0.656 x/dd 11.75 (4) tp/d = l.32 + l.24 (x/d) for 1.35 6 x/d
- 13.5 (5)
O where is = concrete thickne,s required to prevent scabbing tp = concrete thickness required to prevent parforation Equations (4) and (5) were later extrapolated for small values of x/d (Reference
- 20) giving, ts/d = 7.91 (x/d) - 5.06 (x/d)2 for x/df:: 0.65 (6) tp/d = 3.19 (x/d) - 0.718 (x/d)2 for x/d61.35 (7)
A 10 percent margin on thickness has beera applied in the use of equations (6) thru (7) as recommended in Reference 20. O DC-84-13 28
Limited penetration and scabbing was predicted for the set of bounding heavy [y load drops considered; however, in no case were the concrete slobs predicted to be perforated. t OVERALL RESULTS AND CONCLUSIONS 1 Table 2 provides the overall results of the systems evaluations. The term "OK" is used where the systems evaluations demonstrated sufficient redundancy and separation of equipment to assure the capability to perform the required safety function. As indicated, the capability to perform spent fuel cooling, extended , core cooling and safe shutdown, is maintained for postulated load drops in most
. regions.
The term " Potential Problem" is used to indicate those situations for which additional evoluotions and/or restrictions were necessary to assure the capability l to perform the required safety functions. These situations are discussed further in the sections below. A 1. Regions 2,7, and 20 k) The potential problems in the above regions all involve the loss of spent fuel cooling. In Regions 2 and 7 the loss of cooling capability con be limited to the containment fuel pool by operable isolation valves thereby leaving the Fuel i! Building spent fuel cooling h tact. For load drops in Region 20 there is the potential for loss of spent fuel cooling to both the Fuel and Reactor Building pools. i The overall result of these system losses was judged to be acceptable, however, l/ based on the following rationale, if both FPCCU and RHR cooling to the
- Reactor Building fuel pool and/or FPCCU cooling to the Fuel Building spent fuel pool were lost, hoses could be routed to the pools to provide makeup from any availabis voter source. Therefore, spent fuel in the pools would always remain covered with water. Makeup would only be necessary if boiloff from the pool were to occur. Whether or not boiloff would occur, and when, would be highly
' dependent on the amount, power history, and decoy history of spent fuel in the
>I
- x DC-84-13 29 i
h storage rocks; however, even under the most limiting heat load and water level conditions several hours would be available to provide on alternative water O source. As long as makeup con be provided to the pool, there will be no spent fuel domoge and, therefore, no off-site dose consequences of significance. For this reason, it is judged that there is reasonable assurance that Safety Function I, spent fuel cooling, can be accomplished following all postulated load drops. =
- 2. R_egions 13 and 14 in a limited portion of the load handling areas of these regions, there is the potential for o dropped load to damage RHR piping that is nonisoloble from the suppression pool. Since the subject piping is in proximity to the valves whose maintenance may involve use of the monorail it is ryot realistic to restict the load path to satisfy this concern. The potential for opening a nonisolable vent path from the suppression pool to the Auxiliary Building is a concern when reactor coolant discharge (depressurization) to the suppression pool is necessary.
Since such discharge is not necessary when the plant is in Cold Shutdown or Refueling, load handling operations should be restricted to those Operational Conditions. The potential effects of a load drop in these regions are acceptoble O with respect to the Extended Core Cooling and Spent Fuel Cooling safety functions. Based on the above, heavy load handling in Regions 13 and 14 will be restricted to when the plant is in Cold Shutdown or Refueling to resolve this potential problem.
- 3. Region 21 Based on the systems evaluation in Region 21 the consequences of a cask drop would be unacceptable due to the potential for damaging both Divisions of I Standby Service Water. However, we have performed a cask drop analysis for this Region. The results of that analysis indicate that the effects of a cask drop in the washdown area (area obove the tunnel with safe shutdown equipment) are limited to localized spalling of the concrete which will be contained by metal decking installed below the washdown area floor. The structural analysis requires that the cask be carried at a height (bottom of cask) no greater than 6 inches above Fuel Building Elevation i13'-0". With this carry height limit imposed the results of a heavy load drop in this area are acceptable. l l
o DC-84-13 30
-4 a
h
~
- 4. Additional Administrative Requirements y
' In addition to the heavy load handling requirements discussed in Sections 2 and 3 obove, lift height restrictions for loads handled by the Auxiliary Building Tunnel
. Plug Monorail were required as part of the Region 16 systems evaluation. A structural evoluotion was performed to determine limits necessary to assure that a postulated load drop would not cause structural damage to the 95' elevation floor - Based on that evoluotion loods up to 2 tons may be carried no greater than 4
12 feet and loads up to 6 fons no greater than 5 feet above the 95' elevation floor. All limits necessary as a result of the heavy loads handling evaluation will be imposed administratively by procedures. 1 o i. n v DC-84-13 31
I:EFERENCES
\ /
- l. Building Code Requirements for Reinforced Concrete, ACI 318-77, American Concrete .nstitute, December 1977.
- 2. ACI 349-76, Code Requirements for Nuclear Safety-Related Concrete Structures, Appendix C "Special Provisions for impulse and Impactive Effects", American Concrete Institute,1976.
- 3. Kennedy, R. P., "A Review of Procedures for the Analysis and Design of Concrete Structures to Resist Missile impact Effects", Journal of Nuclear Enaineerino and Design, Vol. 37, No. 2, May 1976.
- 4. Mattock, A. H., " Rotational Capacity of Hinging Region in Reinforced Concrete Beams", Flexural Mechanics of Reinforced Concrete, ASCE 1965-50 (ACI SP-12), American Society of Civil Engineers,1965.
- 5. Corley, W. G., " Rotational Capacity of Reinforced Concrete Beams",
Journal of Structural Division, ASCE, Vol. 92, No. ST5, Proc. Paper 4939,
'~'
Oct.1976, pp.121-146.
- 6. " Design of Structures for Missile impact", Topical Report BC-TOP-9A, Bechtel Power Corporation, September 1974.
- 7. Structures to Resist the Effects of Accidental Explosions, TM5-1300, Department of the Army, Washington, D.C., July 1965.
- 8. Neville, A. M., Properties of Concrete, J. Wiley & Sons, New York,1975.
- 9. Design of Structures to Resist the Effects of Atomic Weapons - Strenoth cf Materials and Structural Elements, TM5-856-2, Department of the Army,
'#ashington, D. C., August 1965.
i p) DC-84-13 32
, . _ . _ _ , . . _ . _ _ , _ __ , _ _ - . _ _ - _ . _ , . . ~ . , , _ . _ _ _ _ - - . _ _ _ _ _ _ - - _ - _ - - _ . - _ , . . - _ _ _ . . . _ . - - _ _ . _ _
- 10. " Evaluation of Heavy Load Handling Operations at River Bend Station; (4)-
v Report. No. 1, Interim Actions and General Guidelines", TERA Corporation, April,1984. II. Wang, C. K. and Salmon, C. G., Reinforced Concrete Desion, Intext Educational Publishers, New York,1973.
- 12. Ferguson, P. M., Reinforced Concrete Fundamentals, J. Wiley, New York, 1973.
- 13. Untrover, R. E. and C. P. Siess, " Strength and Behavior in Flexure of Deep Reinforced Concrete Beams Under Static and Dynamic Loading," Civil Engineering Studies Structural Research Series Report No. 230, University of Illinois, Urbano, October 1961.
- 14. Austin, W. J., et of, "An Investigation of the Behavior of Deep Members of Reinforced Concrete and Steel," Civil Engineering Studies Structural Research Series No.187, University of Illinois, Urbana, January 1960.
IS. de Paiva, H.A.R., ono C. P. Siess, " Strength and Behavior in Shear of Deep Reinforced Concrete Beams Under Static and Dynamic Loading," Civil Engineering Studies Structural Research Series Report No. 231, University of Illinois, Urbana, Oct.1961.
- 16. de Paiva, H.A.R., and W. J. Austin, " Behavior and Design of Deep Structural Members - Part 3 - Tests of Reinforced Concrete Deep Beams," Civil Engineering Studies Structural Research Series No.1974, University of Illinois, Urbana, March 1960.
- 17. Winemiller, J. R. and W. J. Austin, " Behavior ar.J Design of Deep Structural Members - Part 2 - Tests of Reinforced Concrete Deep Members with Web and Compression Reinforcement," Civil Engineering Studies Structural Resecrch Series Report No.193, University of Illinois, Urbano, August 1960.
O DC-84-13 33
18.. Newmark,- N. ' M. and J. D. Haltiwanger, " Air Force Design Manual -- Principles and Practices for Design of Hardened Structures," AFSWC-TDR-62-138, December 1962.-
- 19. Crawford, R. E., et al, "The Air Force Manual for Design and Analysis of Hardened Structures," AFWL-TR-74-102, October 1974.
- 20. Civil Engineering and Nuclear Power, Report of the ASCE Committee on impactive and impulsive Loads, Vol. V, American Society of Civil Engi-neers, September 1980.
' 21. Structural Analysis and Design of Nuclear Plant Facilities, American Society of Civil Engineers,1980.
- 22. Vassallo, F. A., Missile impact Testing of Reinforced Concrete Panels, HC-5609-D-1, Calspan Corporation, January 197S.
- 23. Stephenson, A. E., " Full Scale Tornado Missile impact Tests," Electric Power Research Institute, Final Report NP-440, July 1977.
- 24. Beth, R. A. and Stipe, J. G., " Penetration and Explosion Tests on Concrete Slabs", CPPAB Interim Report No. 20, January 1943.
- 25. Beth, R. A., " Concrete Penetration" OSRD-48S6, National Defense Research Committee Report A-319, March 1945.
- 26. " Structural Analysis of River Bend Pressure Vessel Head Drop, Shroud Head Assembly Drop, and Steam Dryer Assembly Drop Conditions," MAR 84-22, Rev.1, July 1984.
[
- 27. " Load Drop Analysis", Stone & Webster Corp., Calc. No. 201.120.139, May 27,1984. ~
- 28. " Spent Fuel Cask Drop Analysis," Stone & Webster Corp., Cole. No. 12210 (C62.S00), November 8,1982.
DC-84-13 34
TABLEI LOAD DROP REGIONS \ ,) t REGION DESCRIPTION HANDLING SYSTEMS I Reactor vessel and area above Polar Crone reactor vessel. RB Elevations 70' through 186'3". 2 Upper contoinment fuel storage Polar Crane pool and dryer storage creo. RB Elevations 146' through I86'3". 3 Reactor building hoist area Polar Crone including level below hoist area floor. RB Elevations 70' through 186'3". 4 Reactor building - east side. Polar Crone RB Elevations 162'3" through I86'3". I 5 Separator storage area (north). Polar Crane [* RB Elevations 162'3" through I86'3". 6 Separator storage area (south). Polar Crane RB Elevations 137'3" through 186'3". 7 Reactor building - west side. Polar Crane RB Elevations 162'3" through I86'3". 8 Containment drywell - east side. Drywell MSIV and RB Elevations 70' to 154'7". Relief Volve Monorail 9 Containment drywell - west side. Drywell MSIV and
-RB Elevations 70' to 154'7". Relief Volve Monoroll 10 Main Steam line space. AB MSIV Monorails Elevations i14' to 161'3".
II RHR B transfer spaces. AB RHR B&C Pump Elevations 70' to 134'6" Monorail b) v
TABLE I (Continued) 4
?
w) . RECl0N !~GCal ' TlON HANDLING SYSTEMS l 12 RHR A transfer spaces. AB RHR A Pump elevatiora 70' to 134'6". Monorail 13 Mainsteem/Feedwater piping MSIV and FWlV orea - east side. AB Monorail Elevations 95'9" to 133'. FW Valve Hoist
'14 Mainsteam/Feedwater piping MSIV and FWlV orea - west side. AB Monorail Elevations 95'9" to 133'. FW Volve Hoist 15 RHR C transfer space. AB RHR B&C Pump Elevations 70' to 105'6". Monorail 16 Auxiliary Building Tunnel Aux. Bldg. Tunnel access region. AB Elevations Plug Monorail 70' to 108'.
17 Control Building equipment Holst Area
' handling region. CB Elevations Monorails e
70' to 156'. 18 Control Building Tunnel access Floor plug Monorail region. CB Elevations 70' to l 15'. 19 Spent fuel storage pool- Fuel Building Bridge west end. FB Elevations 70' Crane to 138'. 20 Fuel Building region west Fuel Building Bridge
' of SFP. FB Elevations 70' Crane to 138'.
21 Cask handling arco. FB Spent Fuel Cask Elevations 70' to 151'. Trolley I w , l t
i TABLE 2 OVERALL RESULTS
~
Safety 1 2 3 Function Spent Fuel Extended Safe Resolution Reg, ion Cooling Core Cooling Shutdown (Section f) 2 Potential OK N/A I Problem 3 OK OK N/A ! 4 OK OK N/A Reactor 5 OK OK N/A Building 6 OK OK N/A 7 Potential OK N/A l Problem 8 OK OK N/A 9 OK OK N/A 10 OK OK OK II OK OK OK 12 OK OK OK Auxiliary 13 OK -OK Potential 2 Building Problem 14 OK OK Potential 2 Problem 15 OK OK OK 16 OK OK OK 4* Control 17 OK OK OK Building 18 OK OK OK 1 20 Potential OK OK l Fuel Problem Building 21 Potential Potential Potential 3 Problem Problem Problem O
- Discusses necessary lift height restrictions
_ _ _ _ _ _ - _ - _ __ ______ ______ . _. . _ _ _ ~ _ . . . O O O e6 ? ?? ? ? 2xP ... . . . . .. .
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- RIVER BEFO STATION l FEAVY LOAD DROP IMPACT REGIONS l RFACTOR Rt MLDING/At JXILIARY BtMLDING. EL 95'-9"
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J FIGURE 3 i RIVER BEm STATION 1-EAVY LOAD DROP IMPACT REGIONS REACTOR BUILDING / AUXILIARY BUILDING, EL i 14'-0"
d [ [ /. o s : l i t e l i r l l l l ' A 0" A 3' R- 0 R 1'- O 58' 1 O6 l N1 N N1 OL ME O OL I C ME E 0"-
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