ML20082H107
| ML20082H107 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/31/1983 |
| From: | STONE & WEBSTER ENGINEERING CORP. |
| To: | |
| Shared Package | |
| ML20082H096 | List: |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR NUDOCS 8312010029 | |
| Download: ML20082H107 (470) | |
Text
{{#Wiki_filter:i J.O.No. 14235.17 NUREG-0612 NINE-MONTH REPORT t CONTROL OF HEAVY LOADS AT NUCLEAR POWER PLANTS CRYSTAL RIVER UNIT 3 FLORIDA POWER CORPORATION ST. PETERSBURG, FL r-) / ' Q l l l l l l October 1983 i Stone & Webster Engineering Corporation Cherry Hill Operations Center Cherry Hill, New Jersey 0312010029 831123 DR ADOCK 05000302 PDR
TABLE OF CONTENTS [}
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Title Page 1.0 PURPOSE i 3 2.0
SUMMARY
1 2.1 General 1 2.2 Referenced Documents 1 2.3 Method of Analysis 2 2.4 Structural Criteria 3 2.5 Mechanical and Electrical Criteria 5 3.0 RESPONSE TO ENCLOSURE 3 OF NRC LETTER DATED DECEMBER 22, 1980 8 (The following section numbers are from the above-referenced document.) 2.2 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN THE VICINITY OF FUEL STORAGE POOLS 9 2.3 SPECIFIC REQUIREMENTS OF OVERHEAD HANDLING SYSTEMS OPERATING IN THE CONTAINMENT 12 fN 2.4 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN PLANT AREAS CONTAINING EQUIPMENT REQUIRED FOR REACTOR SHUTDOWN, CORE DECAY HEAT REMOVAL, OR SPENT FUEL POOL COOLING 15
4.0 CONCLUSION
S AND RECOMMENDATIONS 18 APPENDIXES Appendix A Hazard Elimination Tables Appendix B Load Drop Envelopes Appendix C Plant Drawings ()
TABLE OF CONTENTS (CONT) ()~ Appendixes (Cont) Page Appendix D Decision Chart Appendix E Information Requested in Attachment 1 to Enclosure 3 of NRC Letter dated December 22, 1980, " Single Failure-Proof Handling Systems" (to be provided by October 31, 1984) Appendix F B&W Reactor Vessel Head Drop Analysis Appendix G Calculations
- 1. Load Drop Analysis for RCCR-1
- 2. Load Drop Analysis for CWCR-1
- 3. Load Drop Analysis for SFHT-7
- 4. Load Drop Analysis for RCCR-2
- 5. Load Drop Analysis for FHCR-7
- 6. Head and Internals Handling Fixture Assembly.
- 7. Postulated Spent Fuel Pool Gate Drop into Fuel Pool
() 8. Postulated Drop of 7,000-lb Jack onto Spent Fuel Fool Missile Shields
- 9. Auxiliary Building Crane Girder Buckling Analysis
- 10. Reactor Head Drop with Rotation from Top of Fueling Cavity Appendix H Proposed Technical Specification Changes Appendix.I Core Flood Piping Analysis l
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s / 1.0.0 PURPOSE 1.1.0 D. G. Eisenhut's letter (Reference 1) transmitted fs NUREG-0612 dated July-1980, Control of-Heavy Loads ! (_) at Nuclear Power Plants, and required utilities to review their overhead load handling equipment, systems, and procedures to preclude the possibility of a load drop accident, the consequences of which ~ > could affect the safety of the plant. The purpose of this report is to evaluate the cranes and monorails at the Crystal River Unit 3 Nuclear Generating Plant (CR-3) in accordance with the criteria of NUREG-0612, its attachments, and a USNRC letter dated December 22, 1980. The evaluation is intended to determine system compliance or noncompliance with the criteria of NUREG-0612. Based upon this evaluation, recommendations to upgrade the safety of heavy load handling at CR-3 are made. t O O i
m i 2.0.0
SUMMARY
2.1.0 GENERAL 2.1.1 The following sections describe the specific ; criteria and approach utilized by the structural, mechanical, electrical, and nuclear technology di'sciplines in evaluating the load handling systems at CR-3. 2.2.0 REFERENCED DOCUMENTS 2.2.1 The following referenced documents were used as the basis for this report:
- 1. USNRC letter signed by D. G. Eisenhut dated December 22, 1980, with enclosures:
- a. NUREG-0612, Control of Heavy Loads at Nuclear Power Plants.
- b. Staff Position -
Interim Actions for Control of Heavy Loads.
- c. Request for Information on Control of
, Heavy Loads.
i' 2. USNRC letter signed by D. G. Eisenhut dated {) February 3, 1981.
- 3. FPC Responses to Section 2.1 (of Enclosure 3 to USNRC letter dated December 22, 1980) dated September 2, 1981.
l l f l l l O 1 l
- 4. NRC comments to Ssetion 2.1 dated January 20, 1982.
() 5. FPC responses Nos. 3F-0682-17 to NRC and comments; 3F-1282-02, FPC Letter dated June 15, 1982, and December 1, 1982, respectively.
- 6. CR-3 FSAR, Revision 3, dated July 1, 1983 (includes changes through December 31, 1982).
- 7. USNRC Regulatory Guide 1.29, Seismic Design Classification, Revision 3, dated September 1978.
- 8. NUS Report 3874, 1981. The Control of Heavy Loads. at Crystal River 3. NUREG-0612, Six-Month Report.
2.3.0 METHOD OF ANALYSIS 2.3.1 A heavy load is defined by NUREG-0612 as any load that weighs more than the combined weight of a single spent fuel bundle and its handling tool, which is 2,750 lb for CR-3. 2.3.2 The consequences of a heavy load drop were evaluated for the worst case of all identifiable O loads carried by the following cranes:
- 1. Reactor Building Polar Crane (RCCR-1).
- a. 180T main hook
- b. 30T auxiliary hook
- 2. Auxiliary Building Crane (FHCR-5).
i ! a. 120T main hook l
- b. 15T auxiliary hook
- 3. Reactor Vessel Tool Handling Jib Crane (RCCR-2) (Reactor Building).
- 4. Reactor Building Mechanized Scaffold (Reactor Building roof).
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- 5. Spent Fuel Pool Missile Shield Crane (FHCR-7)
(Auxiliary Building).
- 6. Spent Fuel Pool Gate Hoist (SFHT-7) (Auxiliary Building).
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- 7. Intake Gantry Crane (CWCR-1).
2.3.3 The evaluation is intended to determine compliance f'/\, x_ or noncompliance with the following criteria (NUREG-0612, Section 5.1):
- 1. Any release of radioactive material that may result from damage to spent fuel based on calculations involving accidental drcpping of a postulated heavy load will produce doses that are less than or equal to one-fourth the limits of 10 CFR Part 100 (i.e., 2 75 rem thyroid, 2 6.25 rem whole body).
- 2. Damage to fuel and fuel storage racks, based on calculations involving accidental dropping of a postulated heavy load, will not result in a configuration of the fuel such that K epp is, ;
larger than 0.95. ; 2
- 3. Damage to the reactor vessel or the spent fuel l pool, based on calculations of damage following accidental dropping of a postulated heavy load, is limited so as not to result in )
water leakage that could uncover the fuel. (Makeup water provided to overcome leakage should be from a borated source of adequate
) concentration if the water lost is borated.)
- 4. Damage to equipment in redundant or dual safe shutdown paths, based on calculations assuming the accidental dropping of a postulated heavy load, will be limited so as not to result in loss of required safe shutdown functions.
2.3.4 A drop envelope was developed for each crane based on the limits of hook travel and lifted loads. The drop envelope was then projected down through the structure; any critical components (piping, wiring, i equipment, reactor vessel, spent fuel pool, as l defined by NUREG-0612) were located and identified. l A decision chart (Appendix D) was then used as a guide in determining if the effects of a heavy load drop comply with the criteria of NUREG-0612. 2.4.0 STRUCTURAL CRITERIA 2.4.1 An impact analysis, within the drop envelope, was performed to determine the survival af a critical system or component as defined by UUREG-0612. 2.
4.2 REFERENCES
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Soo Appandix G for list of references. 2.4.3 ASSUMPTIONS
- 1. Initial velocity of load is zero.
(Conservative because maximum height is used.)
- 2. The load strikes the target such that the maximum load is imparted to the surface.
- 3. Any intermediate targets are ignored. Primary target takes full impact.
- 4. No energy dissipation by crushing of the missile is taken credit for.
- 5. The load may be dropped at any location in the crane travel area except. where physical interference is present.
- 6. If drag forces are present, they may be considered.
- 7. The least energy failure mechanism, based on yield line analysis, is considered. All other failure mechanisms are less critical, unless otherwise noted.
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- 8. Analysis is based on a bilinear elastic perfectly plastic curve that represents the stress / strain relationship, unless otherwise stated.
- 9. The effects of existing loads and deflections are ignored for the initial drop analysis. If
. the target does not fail, and a more accurate analysis is required, then the effects of existing loads and deflections will be considered to determine their effe'ct.
- 10. Compression steel is ignored in determining ultimate moment capacity.
- 11. Concrete slab and beam boundary conditions are based on available information and conservative assumptions.
- 12. Concrete slabs and beams are assumed to deform plastically at failure and separate into segments at the yield lines.
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- 13. Failure is considered when the ductility ratio exceeds that as stated in Reference 2, unless otherwise noted.
- 14. If the energy absorbtion analysis for concrete indicates that the target is acceptable, a localized failure check is made.
- 15. Columns are first checked assuming Kl/r s 22.
-16. Bucking is not considered for column members with Kl/r.s 22.
- 17. Passing- columns are again checked with length effects considered.
- 18. Initial deflections to be considered with load drop deflection are dead load and any permanently fixed load such as equipment.
{ Earthquake, creep, etc, deflections are neglected.
- 19. A combination of structures may be present to resist the load drop; however, to simplify analysis, only on- structure is analyzed.
- 20. When more than one block assembly exists (i.e., main and auxiliary hook) the more t critical load is incorporated into the analysis.
- 21. Structural steel supporting concrete slabs is not included in the development of barrier resistance.
- 22. FPC/CR3 FSAR Section 5.4.5.1 states that any vertical seismic response is assumed to be insignificant. Therefore, no induced accelerations other than gravity are considered.
- 23. The ratio 9t = E g /E, is assumed to remain constant during impact and resistance; any increase in E due to a dynamic increase factor is assumed to be negligible.
2.4.4 For method of analysis see Appendix G. 2.5.0 MECHANICAL AND ELECTRICAL CRITERIA 2.5.1 The- critical station mechanical and electrical systems / components were identified utilizing the criteria set forth in NUREG-0612 and Reference 7.
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Pursuant- to NUREG-0612, post-accident systems were not evaluated unless such systems serve a dual function in normal station operation. 3 O Such systems are exempt from analysis since a load drop accident is not assumed to occur in combination with other postulated accidents (i.e., pipe rupture at power, etc). 2.5.2 The evaluation considered critical systems that are required to achieve and maintain a safe shutdown condition. It also considered systems which are required to maintain decay heat removal during refueling operations. Systems .or portions of systems which were evaluated are: , 1. Reactor coolant.
- 2. Spent fuel ~ pool cooling.
*3. Segments of the nuclear service and decay heat sea water, nuclear services closed cycle cooling and decay heat closed cycle cooling required for:
- a. Decay heat removal from the reactor.
- b. Cooling the spent fuel storage pool.
- 4. Portions of the steam and feedwater systems (in excess of 2 1/2 in. in diameter) extending from and including the secondary side of steam generators up to and including the outermost containment isolation valves,- up to and including the first valve (including a safety or relief valve) that is either normally closed or capable of automatic closure during all modes of normal reactor operation.
- 5. Systems or portions of systems that are required to' supply fuel to emergency diesel generators.
2.5.3 Compliance with NUREG-0612 will be based on demonstrating:
- 1. Survival of required systems after load drop to attain safe shutdown condition from modes 3-6. There are no heavy loads lifted in the primary containment during normal
- Systems considered for decay heat removal during refueling operation.
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L operation. Several loads are lifted during mode 3, hot standby. Most heavy loads are lifted, however, during modes 4, 5, and 6.
- 2. Survival of required systems after load drop during refueling operation to provide for decay heat removal from the core.
2.5.4 The B&W Reactor Vessel (RV) head analysis, Appendix F, discusses the consequences of dropping the RV head on the vessel. 7
. . _ . . --_ . _ _ . - - _. - . _ _ . . - _ _ ~ - - ._. - . _ . _ _ _ _ _ _ _ _ . _ _ _ . . _ _ _ _. . . _ _
1 t 1 T i i ? 1 f } r i 4 f 6. ! i i j' Section 3.0 RESPONSE TO ENCLOSURE 3 OF NRC LETTER DATED DECEMBER 22, 1980
' ~
l Sections 2.2, 2.3, and 2.4 l-l l l P b I
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2.2 SPECIFIC REQUIREMENTS- FOR OVERHEAD HANDLING SYSTEMS OPERATING IN THE VICINITY OF FUEL STORAGE POOLS
) NUREG-0612, Section 5.1.2, provides guidelines concern-ing the design and operation of load-handling 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 likelih'ood of a load drop which might damage spent fuel is extremely small, or that the estimated consequences of such a drop will not exceed the limits set by the evaluation criteria of NUREG-0612, Section'5.1, Criteria I through III.
- 1. Identify by name, type, capacity, and equipment designator any cranes physically capable (i.e.,
ignoring interlocks, movable mechanical stops, or operating procedures) of carrying loads which could, if dropped, land or fall into the spent fuel pool.
RESPONSE
Equipment Name Tyne Capacity Designator a) Auxiliary Overhead 120/15 tons FHCR-5 Building Bridge Crane b} FHCR-7 b) Spent Single 10 tons Fuel Pool Leg Missile Gantry l Shield ' l Crane l c) Spent Chain Hoist 3 tons SFHT-7 Fuel Pool Suspended Gate Hoist from a Monorail d) Reactor Specialty 5 tons - Building Mechanical Scaffold
- 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 load can fall into the fuel-storage pool.
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RESPONSE: No exclusions beyond those identified in the 6-Month Report. " (^l x_/
- 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 crane so evaluated, provide the load-handling-system (i.e., crane-load-combination) information specified in Attachment 1.
RESPONSE: The Auxiliary Building Crane will be modified to meet the criteria of NUREG-0554. (See Appendix E for the information requested by Attachment 1 of Enclosure 3 to USNRC letter dated December 22, 1980.)
- 4. For cranes identified in 2.2-1 above, not categorized according to 2.2-3, demonstrate that the criteria of NUREG-0612, Section 5.1, are satisfied. Compliance with Criterion IV will be demonstrated in response to Section 2.4 of this request. With respect to Criteria I through III, provide a discussion of your evaluation of crane O operation in the spent fuel area determination of compliance. This response should include the following information for each crane:
and your
- a. Which alternatives (e.g., 2, 3, or 4) from those identified in NUREG-0612, Section 5.1.2 have been selected.
RESPONSE: The responses in Appendixes A-6 and A-7 are
- based on Section 5.1.2-4 from NUREG-0612.
- 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 procedures invoked to ensure proper authorization of bypass or removal, and provide any related or proposed technical specification (operational and surveillance) provided to ensure the operability of such electrical interlocks or mechanical stops. RESPONSE: Not applicable.
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- c. Whnre reliance is placed on crano operational i limitations with respect to the time of the storage of certain quantities of spent fuel at !
/~) specific post-irradiation decay times, provide '
~/ present and/or proposed technical specifications and discuss administrative or i physical controls provided to ensure that j these assumptions remain valid.
RESPONSE: See Appendix H.
- d. Where reliance is placed on the physical location of specific fuel modules at certain post-irradiation decay times, provide present i and/or proposed technical specifications and discuss administrative or physical controls provided to ensure that these assumptions remain valid.
RESPONSE: See Appendix H.
- e. Analyses performed to demonstrate compliance with Criteria I through III 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
-)
qj 4, as performed. appropriate, for each analysis RESPONSE: The only heavy load lifted by the Spent Fuel Pool Missile Shield Crane, EHCR-7, is the ~ ' missile shields. Since the missile shields float, Criteria I through III are satisfied. The Spent Fuel Pool Gate Hoist, SFHT-7, lifts the gate, which separates the A-B spent fuel pools, to its stored position in a slot along the A pool wall. Dropping the gate could result 'in the gate hitting either the fuel pool floor or the spent fuel racks. If the gate hits the floor the liner will be penetrated; however, the concrete will not fail. This could result.in a small amount of leakage which would be compensated by the borated makeup water capability of the spent j fuel cooling system. If the gate hits the racks, significant rack and fuel assembly damage results, with excessive dose releases possible if damage occurs to spent fuel assemblies that have been stored for <50 days. Therefore, administrative controls will be
- changed to require a 50-day interval between f
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,- -, . - - - . - - -. -- - - - _ _ _ = - - - - . - - . - . - . - . - - , -
storaga of fuel in the pools and movement of the gate. In addition, a requirement shall be added to the procedure such that the spent () fuel pool gate shall not be moved if fuel other than that discharged from the RV at the end of the most recent fuel cycle is stored in the spent fuel pool. Refer to Appendixes G-3 and G-7 for detailed analyses. The Reactor Building Mechanized Scaffold lifts the stud tensioning hydraulic jack above the Auxiliary Building roof. Dropping this j ack will impact the spent fuel pool missile shields. Appendix G-8, Postulated Drop of 7,000-lb Jack onto Spent Fuel Pool Missile Shields, shows that the shields are able to withstand the drop and thereby prevent the jack from falling into the spent fuel pool. 2.3 SPECIFIC REQUIREMENTS OF OVERHEAD HANDLING SYSTEMS OPERATING IN THE CONTAINMENT NUREG-0612, Section 5.1.3, provides guidelines concerning the design and operation of load-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 O likelihood of a load-drop which might damage spent fuel is extremely small, or that the estimated consequences of such a drop will not exceed the limits set by the evaluation criteria of NUREG-0612, Section 5.1, Criteria I through III.
- 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.
RESPONSE
Equipment Name Type Capacity Designator a) Reactor Overhead 180/30 tons RCCR-1 Building Bridge Polar Crane O 12
b) Reactor Monorail 2 1/2 tons RCCR-2 L Vessel Boom , Tool Handling i Jib Crane
- 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 the movement of any load
. .either directly over the reactor vessel or to such a location where in the event of any load-handling-system failure, the load may land in or on the reactor vessel. , RESPONSE: . No exclusions. ~
- 3. Identify any cranes listed in 2.3-1, above, which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely small for all loads to be carried and basis for this evaluation (i.e., complete
- . compliance with NUREG-0612, Section 5.1.6, or i partial compliance supplemented by suitable alternative or additional design features). For each crane so evaluated provide the load-handling-system (i.e., crane-load-combination) l' information specified in Attachment 1.
RESPONSE: Not Applicable.
- 4. For cranes identified in 2.3-1, above, not
! categorized according to 2.3-3, demonstrate that l the evaluation criteria of NUREG-0612, Section 5.1, l are satisfied. Compliance with Criterion IV will L be demonstrated in your response to Section 2.4 of this request. With respect to Criteria I through III, provide a discussion of your evaluation of crane operation in the containment and your determination of compliance. This response should include the following information for each crane:
- a. Where reliance is placed on the installation
! and use of electrical interlocks or mechanical stops, indicate the circumstances under which these protective devices can be removed or j bypassed and the administrative procedures !' invoked to ensure proper authorization of such action. Discuss any related or proposed technical specification concerning the
- bypassing of such interlocks.
RESPONSE: Not Applicable. 13 I i 5
. - . - _ . . , . - . . . . - - - , - . - - . ~ _ _ - - . - . - - .. - . - , . . - - - - - . - - - -
- b. Where reliance is placed on other, site-specific considerations (e.g., refueling sequencing), provide present or proposed Os technical specifications and discuss administrative or physical controls provided to ensure the continued validity of such considerations.
RESPONSE: Not Applicable.
- c. Analyses performed to demonstrate compliance with Criteria I through III should conform with.the guidelines of NUREG-0612, Appendix A.
Justify any exception taken to these guidelines, and provide the specific information requested in Attachments 2, 3, or 4, as appropriate, for each analysis performed. RESPONSE: The reactor vessel head with the lifting ring assembly attached is the load having the greatest potential for damage to the spent reactor core if dropped during refueling operations. The analysis of this load drop on the reactor vessel has been performed by Babcock and Wilcox and is presentad in Appendix F. The maximum vertical lift of the reactor vessel head over the reactor vessel is as indicated in Appendix F. Procedures will be developed to ensure that the reactor vessel head will be moved horizontally away from the reactor vessel at or before the maximum height is reached. A secondary drop has been investigated by SWEC. The reactor vessel head was postulated l , to drop while the head was over the fuel l transfer canal wall. The reactor vessel head , would fall onto the shield wall surrounding the reactor vessel and secondarily hits the reactor vessel as the top of the head rotates down. The total amount of energy imparted to
- the reactor vessel is equal to or less than that considered in the B&W analysis.
Therefore, the B&W drop is considered a worst case. See Appendix G-10 for the detailed analysis. Load drops which are not on or int.o the reactor vessel have no radiological consequences as outlined in evaluation 14
Critoria I through III. Therefore, for load drops which do not impact the reactor vessel, Criteria I through III are satisfied. 2.4 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN PLANT AREAS CONTAINING EQUIPMENT REQUIRED FOR REACTOR SHUTDOWN, CORE DECAY HEAT REMOVAL, OR SPENT FUEL POOL COOLING NUREG-0612, Section 5.1.5, provides guidelines . concerning the design and operation of load-handling systems in the vicinity of equipment or components required for safe reactor shutdown and decay heat removal. Information provided in response to this i 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 l 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 2.1-3c.
- 1. Identify any cranes listed in 2.1-1, above, which you have evaluated as having sufficient design i
O features to make the likelihood of a load drop extremely small for all loads to be carried and the complete basis for this evaluation (i.e., compliance with NUREG-0612, Section 5.1.6, or partial compliance supplemented by suitable alternatives or additional design features). For each crane so evaluated, provide the load-handling-system (i.e., crane-load-combination) information specified in Attachment 1. RESPONSE: The Auxiliary Building Crane will be modified . to meet the criteria of NUREG-0554. (See Appendix E for the information requested by Attachment 1 of Enclosure 3 to USNRC letter dated December 22, 1980.)
- 2. For any cranes identified in 2.1-1 not designated as single-failureproof in 2.4-1, a comprehensive hazard evaluation should be provided which includes the following information:
- a. The presentation in a matrix format of all heavy loads and potential impact areas where damage might occur to safety-related equipment. Heavy loads identification should
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n- ,,..-,,..,-y ,,- .---- - , . , - , , _ _ . , ~ , ,
includo designation and weight or , cross-reference to information provided in , 2.1-3-c. Impact areas should be identified by O construction zones and elevations or by some other method such that the impact area can be located on the plant general arrangement
. drawings. Figure 1 provides a typical matrix.
RESPONSE: Tables A-1 through A-8 of Appendix A and the figures in Appendix B respond to this question in a matrix-format.
- b. For each interaction identified, indicate which of the load and impact area combinations can be eliminated because of separation and redundancy of safety-related equipment,
- mechanical stops and/or electrical interlocks, or other site-specific consideration.
Elimination on the basis of the aforementioned considerations should be supplemented by the
- following specific informationi (1) For load / target combinations eliminated because of separation and redundancy of safety-related equipment, discuss' the basis for determining that load drops will not affect continued system operation (i.e., the ability of. the O system function).
to perform its safety-related (2) Where mechanical stops or electrical interlocks are to be provided, present de' tails showing the areas where crane travel will be prohibited. Additionally, provide a 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 after operations which require bypassing have been completed.
i (3) Where load / target combinations are eliminated on the basis of other, aite-specific considerations (e.g.,
.taintenance sequencing), provide present and/or proposed technical specifications and discuss administrative procedures or physical constraints invoked to ensure i
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,m, _ - . _ _ , - , . , - . _ _ - - . . - . _ . , , , , _ , . , - , - - . .
the continued validity of such considerations. $ /~' RESPONSE: See response to Item 2.a. L y)! All load drops postulated inside the Reactor Building's Containment are postulated to occur at " hot standby" or lower plant conditions. Cranes within the Reactor Building normally are not used during power generating conditions.
- c. For interactions not eliminated by the analysis of 2.4-2-b, above, identify any handling systems for specific loads which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely small and the basis for this evaluation (i.e., complete compliance with NUREG-0612, Section 5.1.6, or partial compliance supplemented by suitable alternative or additional design features).
For each crane so evaluated, provide the load-handling-system (i.e., crane-load-combination) information specified in Attachment 1). RESPONSE: The Auxiliary Building Crane will be modified the criteria of NUREG-0554. O' to meet Appendix E for the information requested by (See Attachment 1.)
- d. For interactions not eliminated in 2.4-2-b or 2.4-2-c, above, demonstrate using appropriate analysis that damage would not preclude operation of sufficient equipment 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 information should be provided:
(1) An indication of whether or not, for the specific load being investigated, the overhead crane-handling system is designed and constructed such that the hoisting system will retain its load in the event of seismic accelerations equivalent to those of a Safe Shutdown Earthquake (SSE). RESPONSE: The only crane which will be analyzed to demonstrate load retention capabilities during 17
an SSE is the Auxiliary Bui,lding Crane, which will be single-failureproof and included in 2.4.1 above. (See -- Appendix E for
-() documentation.)
(2) The basis for any exceptions taken to the analytical guidelines of NUREG-0612, Appendix A. RESPONSE: No exceptions. (3) The information requested in Attachment 4. RESPONSE: Not Applicable. O I O 17A
CE)
4.0 CONCLUSION
S AND RECOMMENDATIONS Based upon the results of the load drop analysis, once the Auxiliary Building Crane is made single-failureproof, operating restrictions are placed on the Spent Fuel Pool Gate Hoist, and the Intake Gantry Crane has had mechanical stops .added to the rail, all cranes and monorails at CR-3 will be in strict compliance with the criteria presented in NUREG-0612. In addition, any temporary or mobile cranes, such as truck cranes and cherry pickers, operating in an area over safety-related equipment will be administratively controlled to preclude operation outside the guidelines of NUREG-0612. O i l I l l I O 18 { l
11 i l 1 I i l l [ I APPENDIX A i N ELIMINATION TABLES l i { e 2 l l l
A LOAD DROP ACCEPTABILITY - HAZARD F-'NATIC',1 TABLE A-1 (REACTOR BUILDINC)
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LC'D DROP ACCEPTABILITY IS BASED UPON THE lu J ELIMINATIC:3 CATECORIES LISTED ON PACE A-9 N Load Path Equipment Title Sketch Impac t ed Equip. Identification No. Load to be Lif ted Load No. Elevation Hazard Elimination Equipment Capacity Load Identification No. Weight 6-Ho. Report and Area Ca tego ry Remarks R.arctor Building Polar Reactor Vessel Missile 26 ton Fig 4-15 180' & 160' - B The potentist exists Cr:ne Shields (3) Reactor Blds , for one of two decay heat supply lines to the RV to RCCR-1 180/30 Ton Pressuriser Missile 15 ton Fig 4-15 & 16 180' & 160' - B be severed. However, only Shields (7) k; actor Bldg one supply line to the RV is required to maintain a Stud Tensioners (2) I ton Fig 4-21 160' & 135' - B safe shutdown condition. Reactor Bldg All other postulated line breaks and/or equipment Stud Handling Tools 170 lb Fig 4-21 160' & 135' - B damage are less critical Reactor Bldg and, therefore, are en-compassed within the above Stud Nos. 15 and 45 640 lb Fig 4-19 160' & 135' - B statement. (No. 15 - North; Reactor Bldg No. 45 - South) Aliment Studs (2) 400 lb Fig 4-19 160' & 135' - B Reactor Bldg ISI Tool - Aires 17 ton Fig 4-24 160' & 135' - B (Assembled) Reactor Bldg East Ladder Cage (Deep 150 lb Fig 4-19 160' & !!8' - B End of Refueling Canal) - Reactor Bldg Crane Main Holst Bottom 5 ton Fig 4-10 & 11 160', 135', 118', & B Block and Hook 95' - Reactor Bldg Crane Auxiliary Holst 800 lb Fig 4-9, 10 & 11 160', 135', !!9 8 , & B Bottom Block and Hook 118' - Reactor Bldg CRDMs and Components 914 lb Fig 4-16 & 22 160' & 135' - B Reactor Bldg Parts and Componento of 2,500 lb Fig 4-24 160' & 135' - B Refueling Hachines Reactor Bldg Fuel Transfer Carriage, 150 lb Fig 4-24 160' & 135' - B Upender, or parts for Reactor Bldg , same Iktch Covers 10 ton Fig 4-16 !!9' - Reactor B Bldg
LOAD DROP ACCEPTABILI'.'Y - HAZARD EI.tMINATION TABLE A--? (REACTOR BUILDING) [ , i ~j LOAD DROP ACCEPTABILITY IS BASED UPON THE
- 1) ELIMINATIC3 CATECORIES LISTED ON PACE A-9 (~~,,1)
Load Path Equipment Title Sketch lapacted Equip. Identification No. Load to be Lif ted Load No. Elevation Hazard Elimination Equipment Capacity Load Identification No. Weight 6-Ho. Report and Area Category Remarks Racetor Building Polar ISI Tool - Artes 16 ton Fig 4-24 160' - Reactor B The potential exists Crane (Unassembled in Crates) Bldg for one of two decay heat supply lines to the RV to be severed. However, only one supply line to the RV
. is required to maintain a safe shutdowo condition.
RCCR-1 180/30 Ton Plenum 58 ton Fig 4-17 135' & 118' - B,E B&W Report - Plenum drcp Reactor Bldg damages fuel; however, the consequences of the drop are such that the core is not uncovered, K 6 aI{kve.95,andradio-release 18 625% of 10CFRIOO limits. This is a wors t-case drop in-to the RV. Internals Storage Stand 8,200 lb Fig 4-24 135' & 118' - B Reactor Bldg Plenum and Core Barrel 162 ton Fig 4-17 135' & 118' - 8,E or Core Barrel Alone Reactor Bldg Specimen Storage Con- 400 lb Fig 4-17 160' & 135' - B tainer (Rack) Reactor Bldg RP Hotors 50 ton Fig 4-18 180', 160', 119', & B The potential exists 95' - Reactor Bldg for the decay heat return line to be severed. If RCP, Coonanents. Struc- 23 ton Fig 4-18 180', 160', 119', & B this occurs, decay heat tural or Supporting Steel 95' - Reactor Bldg supply to the reactor is Above or Around RCPs ultimately via the reactor building sump. Head Lif ting Pendants (3) 850 lb Fig 4-23 160' & 175' - B Reactor Bldg The potential exists for one of two decay heat Head and Internals Handling 6 ton Fig 4-23 160' & 135' - B supply lines to the RV Fixture Potetor Bldg , to be severed. However,
, only one supply line to Reactor Vessel Head with 175 ton Fig 4-17 160' & 135' - B the RV is required to Lif t Rig (Tripod) (Includes Reactor Bldg maintain a safe shutdown Lead Shielding of RV Head) condition.
LOAD DROP ACCEFTABILI1T - HA7_ARD ELIMINATION TABLE A-2 (REACTC3 BillLDINC)
,,~
f ~n LOAD DROP ACCEPTABILI1Y IS BASED UPON TH ( RD ELIMINATICJ CATECC21ES LISTED ON PACE A-9 ! load Path Equipment Title Sketch Impacted Equip. Identification No. Imad to be lifted Imad No. Elevation Hazard Elimination Equipment Capacity load Identification No. Weight 6-Ho. Repo r t and Area Category Remarka
- t rnata Handling Ex- 3.400 lb Fig 4-23 160' - Reactor B tension Bldg Tripod 6 ton Fig 4-23 160' - Reactor u Bldg Inden Fixture and Asso- 12,500 lb Fig 4-23 160' - Reactor B ciated Adapters and Pendanta Bldg and Spreader Ring Stud Support Spacers (in 50 lb Fig 4-20 160' - Reactor B Crate) Bldg a
LOAD DROP ACCEPTABILITY - HAZARD ELIM_INATIC1 TABLE A-3 (REACTOR BUILDINC) LOAD L10P ACCEPTABILITY IS BASED UPON THE n ELIMINATICO CATECORIES LISTED ON PACE A-7 load Path Equipment Title Sketch Impacted Equip. Identification No. Inad to be Lifted load No. Elevation Hazard Elimination Equipment Capacity Load Identification No. Weight 6-Ho. Report and Area Category Remarks Rrctor Building Folar Stud Hold Seal Plugs (58) 200 lb Fig 4-20 160' - Reactor B The potential exista Crani (in Crate) Bldg for one of two decay heat supply lines to the RV to RCCR-1 180/30 Ton CRDH Cooling Water Header 100 lb Fig 4-20 160' & 135' - B be e,evered. However, Spool Pieces Reactor Bldg only one supply line to the RV is required to Service Structure Plat- 1 ton Fig 4-16 160' & 135' - B maintain a safe sisutdown forms (Triangular) Reactor Bldg condition. Refueling Cavity Seal Plate I ton Fig 4-20 160' & 135' - B Reactor Bldg , Fuel Transfer Tube 1 ton Fig 4-19 160' & 118' - B , Covers (2) Reactor Bldg a
LOAD DROP ACCEPTABILITY - HAZARD ELIMINATION TABLE A-4 (REACTOR BUILDINC) LOAD DROP ACCEPTABILITY IS BASED UP03 THE 7 ELIMINATION CATECCilES LISTED ON PACE A-9 load Path i Equipment Title Sketch Impacted
, Equip. Identification No. load to be Lifted load Elevation Hazard E11aination Equipment Capacity load Identification No. Weight and Area Category Remarks Ructor Vessel Tool Unidentified 2 ton 160'. 135'. & B.E The potential exists Handling Jib Crane !!8' - Reactor Blds for one of two decey heat RCCR-2 2 Ton supply lines to the RV to CRDM Cooling Water 100 lb Sheet A-10 160' & 135' - B.E be severed. However, Header Spool Pieces Reactor Blds only one supply line to the RV is required to l
Incore Instrument Tube 300 lb Sheet A-10 160' & 135' - B.E maintain a safe shutdown Plus Tool Reactor Blds condition. Rod Assembly Handling Tool. 150 lb Sheet A-10 160' & 135' - B.E Any drop by RCCR-2 leadscrew Tool. Stator / Reactor Bldg into the RV is less Water Jacket Tool. Hold- severe than the down Bolt Tool plenum drop addressed long Handled Tool and 40 lb Sheet A-10 160' & 135' - B.E ! Attachments Reactor Bldg Source Handling Tool 30 lb Sheet A-10 160' & 135' - B.E Reactor Bids 4 i i i
LOAD DROP ACCEPTABILITY - MAZARD ELIMINATION TABLE A-5 (AUXILIART BUILDINC) LOAD DROP ACCEPTABILITY IS BASED UPON THE (n ) ELIMINATIC2 CATECORIES LISTED ON PACE A-9
%l ,
load Path Equipment Title Sketch Impacted Equip. Identification No. Load to be Lifted Load No. Elevation Hazard Elimination Equipment Capacity Load Identification No. Weight 6-Ho. Report and Area Category Remarks Aaxiliary Building Crane Crane main hoist botton 3-1/2 ton Free movement 162' - Aux Bldg D - FHCR-5 120/15 tons block and hook area Crane auxiliary hoist 700 lb Free movement 162' - Aux Bldg D - bottom block and hook area i New fuel pit missils 7 ton Fig 4-29 162' - Aux Bldg D - shields with SFNS New control component 1,000 lb Fig 4-30 162' - Aux Bldg D - container New control component 132 lb Fig 4-30 162' - Aux Bldg D - Spent fuel cask pit gate 2 ton Fig 4-29 162' - Aux Bldg D - ! Spent fuel pool missile 8,900 lb Fig 4-28 162' - Aux Bldg D - . shields New fuel elevator and 220 lb Fig 4-29 162' - Aux Bldg D - i associated equipment Loading bay hatch cover Less than Fig 4-26 162' - Aux Bldg D - I ton Spent fuel cask (loaded) 28.7 tons Fig 4-26 162' - Aux Bldg D -
& 4-29 New fuel shipping cask 7.300 lb Fig 4-27 162' - Aux Bldg D -
(loaded - 2 assemblies) New fuel assembl!es 1.532 lb Fig 4-30 162' - Aux Blds D - , Missile shield handling 500 lb Free movement 162' - Aux Bldg D - fixtures (SFMS) area Various loads which are Variable. Free movement 162' - Aux Bldg D - hoisted up to the operat- less than area ing deck from the loading 2 ton bay , Various loads on missile Variable, Fig 4-28 162' - Aux Bldg D - shields above spent fuel less than pool "E" af ter transfer- 1,000 lb - l ral by the missile shield crane from the spent fuel pool "A" area
I.OAD DROP ACCEPTABILITY - HA7_ARD ELIMINATION TABLE A-6 (AUXILIART SUILDINC) r r ( v LOAD DROP ACCEPTABILITT IS BASED UPON THE D ELIMINATION CATECORIES LISTED ON PACE A-9 (h) w/ Ioad Path Equipment Title Sketch Iapacted Equip. Identification No. Ioad to be Lif ted load No. Elev.ition Hazard Elimination , Equipment Capacity load Identification Nc. Weight 6-Ho. Report and Area Category Remarks Ructor Building Mechanized Hydraulic Jack 7,000 lb Fig 4-4 209' & 162' - Aux E Calculation No. 14235.17-Sccffold Bldg M-07 5 Ton a
s LOAD DROP ACCEPTABILITY - HAZARD ELIMINATION TABLE A-7 (AUXILIARY BUILDINC) f-s LOAD DROP ACCEPTABILITY IS BASED UPON Tile N ELIMINATICJ CATECC2IES LISTO ON PACE A-9 load Path Equipment Title Sketch Impacted Equip. Identification No. Load to be Lif ted Load No. Elevation Hazard Elimination Equipment Capacity Load Identification No. Weight 6-Mo. Report and Area Category Remarks Spelt Fuel Pool Missile Spent Fuel Pool Missited 6.300 lb Fig 4-29 162' - Aux Didg C The missile shields are Shistd Cantry Crane and Shields A through P with designed to float. Spent Pool Cate Roist SFHS FHCR-7/SFHT-7 Various Items in Spent I ton Fig 4-29 162' - Aux Blds C Missile ahtelds prevent 10/3 Ton Fuel Pit "A" Area (i.e. , these objects from fall-Refueling Machine Parts, ing into the pool. Fuel Transfer Carriages and Hydraulic Units, etc) Spent Fuel Pool "A-8" 3,900 lb Fig 4-29 162' & 118' - Aux C,E The spent fuel pool gate Cate Bldg shall not be moved if fuel other than that discharged from the RV at the eud of the fuel management cycle is stored in the spent fuel pool. A 50-day interval is required between the of floading of spent fuel and movement of the gate. s s
s LOAD DROP ACCEPTABILITY - HAZARD ELIMINATION TABLE A-8 (INTAKE STRUCTURE AREA) ~s I LOAD DROP ACCEPTABILI1Y IS BASED UPON Tile H-L Y' - ELIMINATICQ CATECC21ES LISTED ON PACE A-9 ,%J / Load Path Eq11pment Title Sketch Impacted Eq rip. Identification No. Load to be Lif ted - Load No. Elevation Hazard Elimination _ Equipment Capacity Load Identification No. Weight 6-Ho. Report and Area Category Remarks I taka Centry Crane Intake Structure 50 ton Fig 4-5, 6, 97' - A Hechanical laterlock will CWCR-1 &7 Circulating and be provided to grohibit 50 Tsn Service Water movement of the intake Intake Structure gantry crane over NSRE area. s
A-9 HAZARD ELIMINATION CATEGORIES A- Movement of the load over NSRE or fuel is prohibited by mechanical and/or electrical interlocks. B- System redundancy and separation would allow the plant to be shut down, or maintained in the safe shutdown condition af ter a load d rop . C- The following site-specific considerationn eliminate the need to consider the load drop. o loads handled only when fuel is removed or other condition o Refueling sequence or other condition o loads handled in safe shutdown condition only
. D- Single-failureproof handling system provided in accordance with ,
NUREG-0554. E- Load Drop Analysis () o Drop is acceptable in accordance with the guidelines of para-graph 5.1 of NUREG-0612. o Recommended modifications to be pursued o Structures reinforced o Shock-absorbing materials added i o LLait switches to limit load heights, or load limiters added 1
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RCCR-2 . _. REACTOR VESSEL TOOL HANDLING J1B CRANE FIG. B-2
SPENf' FUEL POOL MISSILE SHJELD CRANE Q v - li SFHT-7 SPENT FUEL POOL GATE HOIST Q' C/ O
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$ TRAVEL OVER N.S.R.E. AREA
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1 178'-5 / 2' BETWEEN BUMPER STOPS _ O l l l CWCR-1 i ~ INTAKE GANTRY CRANE l l O I i FIG. B-4
w.. f A I RAILS j
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AUX. BLDG. l l . DROP ENVELOPE REACTOR BLDG MECHANIZED SCAFFOLD ( R.B. ROOF ) O FIG. B-5
APPENDIX C - PLANT DRAWINGS t FPC Drawing No. L-001-012 Layout, Plan Above Reactor 1.8 Auxiliary and Intermediate 1.9 Building Basement Floor 1.10 El 75' 0" and 95' 0" 1.11 FPC Drawing No. L-001-022 Layout, Plan Above Reactor, 1.13 Auxiliary and Intermediate 1.14 Building Mezzanine Floor 1.15 El 119' 0" 1.16 FPC Drawing No. L-001-023 Layout, Plan Above Reactor, 1.18 Auxiliary and Intermediate 1.19 Building 1.20 El 143' 0" 1.21 FPC Drawing No. L-001-032 Layout, Plan Above Reactor 1.24 Builoing Operating Floor 1.25 El 1,0' 0" and 1.26 Auxiliary Building 1.27 El 162' 0" . 1.28 FPC Drawing No. L-001-042 Layout, Plan Above Reactor 1.31 Building 1.32 El 180' 0" 1.33 FPC Drawing No. L-002-002 Layout, Cross Section Through 1.36 Reactor Building and Auxiliary 1.37 Building 1.38 FPC Drawing No. L-002-003 Layout, Longitudinal Section 1.41 Through Reactor Building and 1.42 Spent Fuel Pit 1.43 FPC Drawing No. CM-303-609 Composite Reactor Building 1.46
- and Intermediate Building 1.47 El 95' 0" 1.48 FPC Drawing No. CM-303-610 Composite Reactor Building 1.51 and Intermediate Building 1.52 El 119' 0" 1.53 FPC Drawing No. CM-303-611 Composite Reactor Building 1.56 and Intermediate Building 1.57 El 162' 0" 1.58 1
v ch-14235.17-1d 08/11/83 155
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a.e seep ta.ble r sm Lasates teemaep h ) a: ere tae the fetts tag as necessary esca that the 3eed drey een be acess= enested (a) Seief erte structores shoes equipseet See G) add shock abeerting esterinie er structores Perfem emelpees se ey. .a (1) (C) add 11 alt eritchee te place 6&e eheve che 43 rep = edtstise teatrict leed height pad land emeld eeJesse F1/4
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- 4. Damage 5512 seek es SER. feet (4) F581 to este ehetaeve med feel t operab peel emellae agesp.
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(C) add 11att evitches to See restrict leed height 88038 Perfere aosJyees se app 11 sable (1
"** ehete the dropped Seef emeld , radiettes
- 1. Selease redsencetwity veteese 71/4
- 2. ammage fee 2 4 font recha 10CFL100 limite Acemptaba
- 3. Demese reester pensel er (2) Feel E 7 Ceeds t tee speet foot peel (3) Wates he.9
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- - - - - - - - - - _ - - - - - 8312010029 -o[.
l l O 1 l l l APPENDIX E The information requested in Attachment 1 to enclosure 3 of NRC letter dated December 22, 1980, " Single Failure - Proof Handling Systems." (To be provided by October 31, 1984) O o
a ,#JJA. .a am.. 4 a u- _.._.A_ _a m .a - - -- , - -- _ _2_ -.-- _ 4A__.. -_,m-_a w-&.._. _, f i i O l 1 I APPENDIX F B&W REACTOR VESSEL HEAD DROP' ANALYSIS 1 t
.a o
l l l
ANALYSIS OF THE EFFECT OF REACTOR Q VESSEL' HEAD DROP DN THE REACTOR VESSEL
- Crystal River Three -
s es 2 e 8 a O M -
-(Hf/
E i!" I' I1 v O 77-1145763-00
$u~s 983 ,
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77-1146763-00 BAW-1805 August 1983
- O i
4
- i ANALYSIS OF THE EFFECT OF REACTOR VESSEL HEAD DROP ON THE REACTOR VESSEL
- Crystal River Three -
t I t
\
i i O l i BABCOCK & WILCOX O Utility Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 Babcock & Wilcox -
'~x .
gd CONTENTS Page
- 1. INTRODUCTION . . . . . . . . . . . . .. . . . . . . . . . .... 1-1
- 2. ANALYSIS OBJECTIVES AND CRITERIA . . .. . .. . . . . . . .... 2-1
- 3. DETERMINATION OF LOADS . . . . . .. . .. . . . . . . . . .... 3-1 3.1. Calculation of Weights . . . . .. . .. .. . .. . .... 3-1 3.2. Calculation of Drop Height .. . . . . . . . . . .. .... 3-3 4 REACTOR VESSEL GEOMETRY AND MATERIALS . . . . . . . .. . .... 4-1
- 5. POINT LOAD ANALYSIS . . . . . . . . . . . . . . . . .. . .... 5-1 5.1. Analytical Method . . . . . . .. .. . . .. . . . .... 5-1
/"'N 5.2. Results . . . . . . . . . . . . . . . . . . . . . . .... 5-3
- 6. UNIFORM LOAD ANALYSIS . . . . .. . . . . .. . . . .. . .... 6-1 6.1. Analytical Method . . . .. . . . . . . . . . . . . .... 6-1 6.2. Results . . . . . . . . . . . . . .. . . . . . . . .... 6-1
- 7. REACTOR COOLANT PIPING STRESSES . . . . . .. . . . .. . .... 7-1
- 8.
SUMMARY
AND CONCLUSIONS . . . . ... . . . . . . . . .. .... 8-1 APPENDIX -- Material Properties and Allowable Stresses . . .... A-1 APPENDIX - References . . . . . . .. . . . . . . . . . . .... B-1 1 List of Figures ' l Figure 3-1. Closure Head With Service Structure and Lifting i Equipment Attached . . . . . . .. . . . . . .. . . . .... 3-4 ) 3-2. Postulated Point Load Case .. . . . . .. .. . . .. .... 3-5 r3 4-1. Longitudinal Section of Reactor Vessel and 4-2 () f 5-1. Support Skirt . . . . . . . .... . . . . . . . . .. .... Three-Dimensional Finite Element Model . . .. . . . . .... 5-5 5-2. Longitudinal Section . . . . . . . . .. . . . . .. . .... 5-6
- 111 - Babcock & Wilcox
p - Figures (Cont'd) t Figure "I* 5-3. Model After Substructuring . . . . .. . . . . . . . . .... 5-7 5-4. Displacement Time History Near Core Flood Nozzle Vertical Component, Point Load Case . . . . . .. .. . .... 5-8 5-5. Displacement Time History Near Core Flood Nozzle Radial Component, Point Load Case . . . . . . .... . .... 5-9 5-6. Transverse Section at Primary Piping . . . . . . . . . .... 5-10 5-7. Support Skirt Detail ... . . . . .. . . . . ... . .... 5-11 6-1. Displacement Time History Near Core Flood Nozzle Radial Component, Uniform Load Case . . . . . . . .. . .... 6-2 6-2. Displacement Time History Near Core Flood Nozzle vertical Component, Uniform Load Case . . . . ... . .. ... 6-3 A-1. Stress-Strain Curve for SA-508 Class 2 at 73F . . .. . .... A-5 A-2. Stress-Strain Curve for SA-516 Grade 70 at 20F .. . . .... A-6 A-3. Stress-Strain Curve for SA-516 Grade 70 at 125F . . . . .... A-7 0 O v
- iv - Babcock 8.Wilcox
- 1. INTRODUCTION A heavy loads analysis sz was performed for the Oconee Nuclear Station to de-termine the effects of dropping heavy loads on on the reactor vessel (RV) as prescrfbed by NUREG-0612.1 A comparative evaluation" of dimensions, materials, weights, and attached piping was conducted to determine the applicability of the Oconee results to the Crystal River Three plant. This report presents the analysis and results from the Oconee work, modified as required to address the Crystal River plant.
The maximum load lifted above the vessel is the RV head and service structure. Therefore, the Oconee analysis was limited to determining the effects of drop-p ping the RV head onto the reactor vessel. The objective was to determine an () allowable height to which the RV head may be lifted above the RV and dropped without violating the requirements of NUREG-0612. The aralysis dealt primarily with a point load in which the head fell at an angle and impacted the RV at a specific point. An earlier study proved that the point load case limits the drop height to a shorter distance than does a drop in which the head impacts the RV uniformly around the flange.2 However, a uniform drop from the reight used in the point load case was also investi-gated in the Oconee analyais. A three-dimensional, half-symmetry, finite element model of the RV and its support skirt was constructed, and the minimum lift height necessary to clear the guide studs (plus a small working margin) was determined. The impact was modeled by lumping the mass of the RV head and its attachments at the top of the RV flange and applying an initial velocity to the flange to represent the impact effects. The analysis was a nonlinear, dynamic transient procedure us-ing the ANSYS finite element computer program.8 The solution was evaluated in p terms of the resultant stresses in accordance with NUREG-0612.1 (.) 1-1 Babcock a Wilcox
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- 2. ANALYSIS OBJECTIVES AND CRITERIA The objective of this analysis is to satisfy the requirements of NUREG-0612 1 as set .forth below. General guidelines for the control of heavy loads are given in Chapter 5.1 of NUREG-0612. The guideline applicable to this analysis is as follows:
Damage to the reactor vessel or the spent fuel pool based on cal-culations of damage following accidental dropping of a postulated heavy load is limited so as not to result in water leakage that could uncover the fuel (makeup water provided to overcome leakage should be from a borated source of adequate concentration if the water being lost is borated). This report presents an analysis performed to satisfy the guideline above. p Regarding the evaluation criteria specified in section 5.1 of NUREG-0612, the V rules of the ASME Boiler and Pressure Vessel Code, Section III, Appendix F, have been selected as an appropriate set of acceptance criteria." Appendix F defines allowable stress limits for level D service conditions. In the ASME Code, level D service conditions are defined as those combinations of conditions associated with extremely low probability postulated events whose consequences are such that the integrity and operability of the system may be impaired to the extent that conditions of public health and safety are in-volved.
- The stress limits of Appendix F are provided for l
l limiting the consequences of the specified event. They are in-I tended to ensure that violation of the pressure retaining boundary will not occur in components or supports which are in compliance with these procedures, t The effects of the heavy loads considered are analyzed to the stress limits l of Appendix F, thus satisfying the requirements of the evaluation criteria in NUREG-0612, Section 5.1.l'" r Appendix A of NUREG-0612 contains guidelines for . conducting an analysis of a ( heavy load drop. Section A-1 contains general guidelines that should be l l 2-1 Babcock & Wilcox
y considered, as appropriate, in any haavy load analysis. These guidelines are I,h Q listed below, followed by a discussicn of how each guideline applicable to the structural evaluation of the reactor vessel is satisfied in this analysis fol-lowing a drop of the RV head. The general guidelines are as follows:
- 1. The load is dropped in an orientation that causes the most severe consequences.
- 2. That fuel impacted is 100 hours subcritical (or whatever minimum is allowed in facility Technical Specifications prior to fuel handling) .
- 3. The load may be dropped at any location in the crane travel area where movement is not restricted by mechanical stops or e.lectrical interlocks.
- 4. Credit may not be taken for spent fuel pool area charcoal filters if hatches, wall, or roof sections are removed during the handling of the heavy load being analyzed, or whenever the building negative pressure rises above (-)1/8 inch (-3 m) water gauge.
n 5. Analyses that rely on the results of Table 2.1-1 or Figures 2.1-1 or 2.1-2 for potential offsite doses or safe decay times should verify that the assumptions of Table 2.1-2 are conservative for the facility under review. X/Q values should be derived from analysis of onsite meteorological measurements based on 5% worst meteorologi-cal conditions.
- 6. Analyses should be based on an elastic-plastic curve representing a true stress-strain relationship.
l
- 7. The analysis should postulate the " maximum damage" that could re-sult, i.e. , we should consider that all energy is absorbed by the structure and/or equipment that is impacted.
- 8. Loads need not be analyzed if their load paths and consequences are scoped by the analysis of some other load.
- 9. To overcome water leakage due to damage from a load drop, credit may be taken for the borated water makeup of adequate concentration re-t l
quired by the Technical Specifications to be available. i 2-2 Babcock s.Wilcox I
N 10. No credit may be taken for equipment to operate to mitigate the ef-facts of the load drop if that equipment is not required by the Technical Specifications to be operable when the load could be dropped. The following is a point-by-point discussion of the consideration of each ap-plicable guideline above in this analysis:
- 1. According to the results of reference 2, the most severe consequences of the drop are associated with the head falling at an angle and im-pacting the vessel flange at a point. A uniform drop was also made from the corresponding height for the point load case.
- 2. Not applicable.
- 3. This analysis is limited to an evaluation of the reactor vessel.
- 4. Not applicable.
- 5. Not applicable.
- 6. True stress-strain curves are used in the analysis (see the appen-dix to this report for material properties).
- 7. The RV head and its attachments are considered rigid and all energy
, is absorbed bf the vessel.
- 8. The RV head with attachments is the maximum load lifted over the vessel.
- 9. Not applicable.
- 10. Not applicable.
Section A-2 of Appendix A to NUREG-0612 contains additional guidelines to be considered when the postulated load drop is the RV head : l
- 1. Impact loads should include the weight of the RV head assembly (in-cluding all appurtenances), the crane load block, and other lifting apparatus (i.e., the strongback for a BWR).
- 2. All potential accident cases occur during refueling operations. As a minimum, the following should be considered:
- a. Fall of the RV head from its maximum height while still on the guide studs followed by impact with the RV flange.
- b. Fall of the RV head from its maximum height considering possible impacting objects, such as the guide studs, the RV flange, the i steam dryer (BWR), or structures beneath the path of travel.
- c. Impact with the refueling cavity wall due to load swings with
( the subsequent drop of the RV head due to failure of the lift-ing device or wire rope. 2-3 Babcock & Wilcox
)
i
- 3. All cases to be considered should be analyzed in the actual medium O present during the postulated accident, e.g. , for a PWR prior to l
l reassembly, the fueling cavity is drained after the head engages the guide studs to allow for visual inspection of the insertion of , control rod drives into the head. During this phase it should be considered that the head will only fall through the air, with none of the drag forces produced by a water environment.
- 4. In those nuclear steam systems (NSSs) where portions of the reactor internals extend above the RV flange, the internals should be anal-yzed for buckling and resultant adverse effects due to the impact loading of the RV head. It should be demonstrated that the energy absorption characteristics (causing buckling failure) of these in-ternals should be such that resultant damage to the core assembly does not cause a condition beyond the acceptance criteria for this analysis.
- 5. Reactor vessel supports should be evaluated for the effects of the transmitted impact loads of the RV head. In the case of PWRs, in
( which the RV is supported at its nozzles, the effects of bending, shear, and circumferential stresses on the nozzles should be ex-amined. For BWRs the effects of these impact loads on the RV sup-port skirt should be examined.
- 6. The RV head assembly should be considered rigid and should not ex-perience deformation during impact with other components or struc-tures.
This analysis is in compliance with these guidelines, as discussed below.
- 1. The weight considered in the analysis includes the RV head and all attachments and the lifting apparatus as described in section 3 of this report.
- 2. a. The drop heights considered in the uniform load analycis exceed the requirements of this guideline,
- b. Since this analysis is concerned with the reactor vessel, pos-sible objects of impact other than the RV are neglected in order to maximize the effects of the load drop on the RV.
m c. See response b above.
'3. All phases of this analysis assume an air medium during the load drop, maximizing impact velocity.
2-4 Babcock 8 Wilcox
4 The scope of this analysis is limited to an evaluation of the reac-O, ' tor vessel, its supports, and the stresses developed in the primary piping.
- 5. The RV support skirt is analyzed for the effect of the load drop.
- 6. The RV head assembly is considered rigid in this analysis.
This concludes the discussion of the analysis guidelines prescribed by NUREG-0612. In addition to these guidelines, the following points are noted:
- 1. The increase in the ultimate tensile strength of steel when subjected impact loading with a high strain rate is conservatively neglected in this analysis.
- 2. The stress in the support skirt due to the dead weight of the in-ternals and water is considered negligible compared to the impact stresses. The dead weight of the RV, however, is included in the
! analysis.
- 3. No structural damping is included. Damping is considered to have a negligible effect on the peak response due to impact loading.
- 4. The stiffness of the primary piping is included, whereas the stiff-ness of the core flood nozzles is considered negligible and thus is
.O not included.
l 2-5 Babcock s.Wilcox
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- 3. DETERMINATION OF LOADS l
3.1. Calculation of Weights ( The load considered in this analysis'is the RV head assembly and all attached ' handling equipment used in lifting it. The head assembly includes the follow-4 ing:
- 1. Reactor vessel closure head (including service structure support flange and control rod mechanism housings).
j 2. Closure head studs, nuts, and washers. 1 3. Service structure. l 4. Control rod drives. l S. Head and service structure fixed pendants.
- 6. Stud parking spacers.
- 7. Chain hoists (four) .
The following handling equipment is uses in lifting the RV head assembly:
- 1. Head and internals handling fixture.
- 2. Internals handling extension.
- 3. Movable pendants (two).
- 4. Slings (three).
- 5. Crane load block. ,
- 6. Rotating strongback.
Figure 3-1 shows the RV head assembly with attached handling equipment. The total weight of the items above is calculated on the following pages. ( O v l i 3-1 Babcock & Wilcox
. Slinas (reference B&W drawing 78407B-00)
, Total length = 73'-6" 2.625in. diameter-A={(2.625)2 = 5.4119 in.2 (5. 9) = 2.76 fe' Volume = (73.5)
Weight = 2.76 (490 PCF) = 1354 lb - use 1400 lb l Internals Extension (reference B&W drawing 105737D-11) Mark 222 (two) - 2 " x 8'-8" x l'-2" Volume = 2 .25 2 (8.67)(1.167) = 3.79 f t 3 Mark 223 (two) - 34" x 14" x 3h" 3 Volume = 2 x 34(14)(3h) 3
= 1.93 ft Mark 327A (two) - use Mark 223 volume Pins (four) - length = 15.25 in., diameter = 6.25 in.
A={(6.25): - 30.7 in.3 . Volume = 4(30.7)(15.25) 3
= 1.08 ft 3 Mark 225 (two) - 11" x 11" x 1h" 3
Volume = 2 x 11 x 11 x 1.5h=0.21ft Mark 227 (one) Volume = 6"(12")(1.5") if3 = 0.06 fe 3 l Total volume of internals extension (ft'): Mark 222 = 3.79 Mark 223 = 1.93 Mark 327A = 1.93 Pins = 1.08 Mark 225 = 0.21 Mark 227 = 0.06 9.0 Weight = (9.0)(490 PCF) = 4410 lb - use 4500 lb 3-2 Babcock & Wilcox
Head Assembly (reference equipment specification 1005162-00~ and reference 49) Weight = 306,960 lb Handling Fixture (reference B&W drawing 167221E-00 and ref. 5) Weight = 13,000 lb Rotating strongback = 800 lb Total = 13,800 ,1b Pendants (two)(reference B&W drawing 133303C-02) Weight = 1400 lb Crane Load Blockwa Weight = 10,000 lb i Total Weight (lb) Head assembly = 306,960 Handling fixture = 13,800 Two pendants = 1,400 Crane load block = 10,000 Three slings = 1,400 Internals extension = 4,500 338,060 - use 343,000 s2 3.2. Calculation of Drop Height ! The drop height used in the analysis is the minimum height necessary to clear the guide studs (3'-3.4375", ref. B&W dwg 149904E-05) plus an amount designat-
- ed as a working margin (l'-8,5625"). This yields a lift height of 5 ft 0 in.
l Because of the geometry of the head, this corresponds to a drop height of 16 in. for the point load case (see Figure 3-2) . l 3- 3 Babcock & Wilcox l
Figure 3-1. Closure Head With Service Stre ure and Lifting Equipment Attached POLAR CRANE HOOK
\
lHTERNALS Re g HANDLING e EXTENSION 10'-9 1/2" e u A 16 HEAD & INTERNALS HANDLING FIXTURE 12'-2 5/8" ASSEMBLY (TRIPOD) TURNBUCKLE PENDANT ' HANDLING FIXTURE ASSEMBLY
/ ,g n
O . 27'-6" M ~ k SL NG SERV CE STRUCTURE REACTOR VESSEL ALlGNMENT STUD k k / -
,;F ' \j 4i.2" 2 1 i
- n O
3-4 Babcock & Wilcox
)
Figure 3-2. Postulated Point Load Case O CENTER OF GRAVITY DIRECTLY TACT POINT CLOSURE HEAD WITH SERVICE STRUCTURE (IN ROTATED POSITION BEFORE VERTICAL OROP) ,, 60" LIFT HEIGHT U (BEFORE ROTATION) 8 q I 3 l i i l
- OP HEIGHT REACTOR l VESSEL I l 1 I I } l
{ O i O ' 3-5 Babcock & Wilcox
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- 4. REACTOR VESSEL GEOMETRY AND MATERIALS The Crystal River Three reactor vessel is skirt-supported. The materials of construction are as follows (reference 51):
Reactor vessel A-508 Class 2 Support skirt SA-516 Grade 70 Lower head SA-533 Grade B, Class 1 The properties for these materials and allowable stresses are tabulated in the appendix. The component dimensions were obtained from the drawings listed as references 7 through 41 and 50. Pertinent dimensions are shown in Figure 4-1. 1 f O t l i O t 4-1 Babcock & Wi!cox
Figure 4-1. Longitudinal Section of Reactor Vessel O and Support Skirt 5/8" _ 2 3/8" _ _ [gi.gn l ' - 2" - r = 8'-7" l'-10 1/4" _ = ,,
= 7'-0" 1'-4 16 ," I /#
l - H \ 1
~
- n - l.?'
~
l SEAL 1.lNE 10'-7 11/I6" 2'-2 7/8" - 6'-10 11/16" 10'-6 3/4" li.0" . 7i-0 3/16" u l q L_ 6 3/16" - 4 1/2" 8 7/16" - = - 7'-1 11/16" 31'-0" I l ' -5 1/2" - 7'-3 1/4"R 6 ' -3 3/8" - 7" 3'-5 1/16"
'l'-4 4 3/8" - 16 1 11 1 g_ 4 h 1, l' l l
u
- 3 7/16" l 7 ' -3 1/4" 6' -7 3/4" g 2" 7'-8 1/4" l L -- U u U g , {
l'-0 1/2" _ ~
!5"
- 9" 2'-1 13/16" -
_-_ _ 7'-5 3/4" _ _ 7" O 4-2 Babcock & Wilcox
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- 5. POINT LOAD ANALYSIS a
5.1. Analytical Method The effect of _ the RV head and attachments impacting the reactor vessel is modeled as an initial velocity problem. The RV head mass is lumped at the top of the RV ledge and is given an initial velocity to simulate the impact. The validity of this approach and the implementation of the method in the ANSYS computer program 3 were verified and documented in reference 2 through test cases run on ANSYS and checked against hand calculations. The loading condition assumes that the failure mechanism would cause the head to fall obliquely to the plane of the RV flange, causing contact over a small i area of the flange. It was further assumed that this contact area is small
~
enough to idealize this condition as a point load. l Referring to Figure 3-1, one type of failure mechanism that could cause point loading would be the failure of one of the three slings attaching the lifting i lugs on the RV head to the head and internals handling fixture, followed by I failure of the other two. If this failure occurred, the head would have an initial rotation about some horizontal axis, which could lead to the RV head position of Figure 3-2. The worst-case vertical point load is with the center of gravity directly over the contact point. The analysis assumed a 16-> inch free fall with the head in a rotated position. The head could actually be raised a minimum of 60 inches (or 5'-0") prior to f ailure before the amount l l of energy transferred to the RV upo 1 impact would exceed that for the 16-inch point load analyzed in this report. l In order to accurately model this system, a three-dimensional,180' (half sye-metry) model of the RV was required (see Figures 5-1 and 5-2). ANSYS was chosen as the finite element computer code because of its nonlinear, dynamic problem-solving capability.3 l I The shell and lower head of the RV were modeled using three-dimensional iso-parametric solid elements. These eight-node elements have three translational 5-1 Babcock & Wilcox _ _ _ _ - . _ _ ~ _
degrees of freedom (DOFs) at each node. They may be redefined as triangular O-
-Q solids and/or tetrahedra as needed. Using triangular plastic shell elements, the support skirt was modeled as a unf. form cylinder 2 inches thick. This three-node element was chosen for its ability to model plastic behavior. It
- has 6 DOFs at each node, three translations, and three rotations. The inter-face of these two element types was accomplished using the Zienkiewicz in-plane rotational stiffness option (as formulated for the shell elements)'.
The stiffness of the attached primary piping was included and represented as three stiffness matrices. These matrices were derived from a separate analy-sis' and require a third element type, the three-dimensional generalized stiff-ness matrix. This element has two coincident nodes, one fixed against any displacement and the other modeling the stiffness of the pipe-to-nozzle in-terface on the reactor vessel. The results from the analysis in reference 2 revealed no undesirable effects
-(plasticity, buckling, etc.) in either the RV shell or the lower head. For this reason, along with the knowledge that the system would behave similarly s for this case, it was decided to substructure the model to increase ics ef-N ficiency. Most of the RV shell was substructured into one large super ele-ment (see Figure 5-3). Sufficient master DOFs were retained to model both the attached piping and the response of the structure. Similarly, most of the lower head was substructured into a second super element.
1 These four element types make up the 180* model. For this model to represent the RV, the translational DOFs normal to the plane of symmetry of nodes were constrained to allow no displacement. This reflective symmetry decreased the complexity of the three-dimensional model while permitting the system to sim-ulate the response of a point load, provided the load is applied in the plane l of symmetry. In addition, the lower nodes of the skirt elements were con-strained in all 6 DOFs, representing the fixity between the support skirt and the concrete floor of the reactor cavity. Reference 2 included a frequency analysis of the model performed to determine
-an appropriate integration time step. ANSYS uses a form of the Houbolt di-l rect integration method to integrate the equations of motion. This method ; has inherent numerical dacping, and the integration time step must be chosen so that the modes of interest are not excessively damped. The results of the i 5-2 Babcock s.Wilcox
frequency analysis showed a fundamental frequency of 64 Hz and included a maxi-() mum frequency of 4010 Hz. The relationship between the integration time step and the aumerical damping is given in reference 3. The response at a particu-lar frequency is damped approximately 1% when the integration time step is 1/30 of the corresponding period of vibration. For this model a time step of 0.00008 second was used to start the transient analysis; later in the analysis a time step of 0.0001 second was used. The model comprises nonlinear elements representing the reactor vessel and two concentrated mass elements at the top nodes to represent the RV head. The nonlinear material properties are given in the appendix. Plastic material behavior is represented by bilinear kinematic hardening. The ANSYS program uses the maximum cetahedral shear stress (or maximum distortion energy) theory of failure 3 , which is in accordance with the requirements of Section III, Appendix F, paragraph F-1321.1 of the ASME Code." This model was then used in a nonlinear dynamic analysis. The loads applied were the weight of the RV applied statically and the combined weight of the RV head and attachments applied dynamically. An initial velocity was applied
\ at the RV ledge to model the impact caused by a 16-fach drop of the RV head in the rotated position.
5.2. Results l The response of the structure due to impact was monitored at nine locations along the plane of symmetry (see Figure 5-2). Representative time history plots of the response are shown in Figures 5-4 and 5-5. Figure 5-4 shows the vertical displacement near the core flood nozzle directly under the load; it reaches a maximum displacement of -0.218 inch (downward) . Figure 5-5 shows the horizontal displacement near the core flood nozzle; it reaches a maximum of 0.254 inch (outward) . 7,eferring to Figure 5-6, the maximum displacements for the primary piping , are as follows: O 5-3 Babcock & Wilcox
Displacement, inches V Nozzle X Y Z Inlet A2 0.220 -0.017 -0.193 Outlet 0.183 0.120 -0.099 Inlet A1 0.181 0.008 -0.067 i Further analysis was performed to obtain stresses in the first 12 elements around the skirt. This represents the segment from 0 to 45' (where O' is directly under the load). Because of the element numbering scheme used, the 12 elements are numbered 1-4, 13-16, and 25-28 (see Figure 5-7). The verti- , cal stresses at the top, middle, and bottom of the elements were obtained, and it is from these stresses that the membrane and bending stress intensities were calculated. The maximum stress intensities were as follows: Stress Allowable intensity, Element No. (appendix), Stress category psi (Figure 5-7) psi Primary membrane 44,375 13 49,000 Primary membrane 44,493 13 73,500 plus bending ' On the basis of these results, the system was checked against possible buck-ling of the support skirt. The approach taken was to use the maximum observed calculated vertical stress component and assume that it acts uniformly around ! the skirt. This " reaction" was then compared to the critical buckling load calculated in the uniform load analysis. The maximum vertical stress (at the bottom of the skirt elements) - 44,402 psi - occurred in element 13. If this value were uniform around the skirt i (with an area of 1115.27 in.2), it would represent a maximum reaction of 8 49.5 x 10 8 lb, which is well below the critical buckling load of 751 x 10 lb (reference 2). O 5-4 Babcock & Wilcox
Figure 5-1. Three-Dimensional Finite Element Model N 7 ~%
/
/ ,
\ _
s y i
~
/
/ '
\ \~
/ c55" % \
jY l \ j / \
/ \
-w
/ / N \.
/
s \ f / % \ .
/ - \
l / j,
]N s \ \ .
# \
INTERIOR VIEW l l l l O 5-5 Babcock & Wilcox _ ,e n - - - - - ,, .m , _ . , . , - - - - - - - - - - ,.e ,- n. ~
Figure 5-2. Longitudinal Section LOAD CLOSURE FLANGE r
~ -
CORE FLOOD N0ZZLES r INLET /0UTLET N0ZZLES r -
,; 7 w -
w - w - i
} I TOP OF SUPPORT = o 7)7 7;T RESPONSE MONITORING LOCATIONS O
5-6 Babcock & Wilcox
Figure 5-3. Model After Substructuring I p _ f RV SHELL SUPERELEMENT l SUPPCRT - SKIRT LOWER HEAD SUPERELEMENT l t l l l l 5-7 Babcock & Wilcox l
Figure 5-4. Displacement Time History Near Core Flood Nozzle Vertical Component, Point Load Case
.160 -
.120 -
s
.080 .
.040 -
U 8e
.000 5 .
5 g .040 . o .
.5 O
cm
.080 -
.120 -
.160 -
.200 .
.240 e i
.0000 .0036 .0071 .0107 .0143 .0179 .0214 .0250 Time, seconds O
5-8 Babcock & Wilcox
3 Figure 5-5. Displacement Time History Near Core Flood Nozzle Radial Component, Point Load Case
.360 -
.320 -
.280 -
g .240 - 5
.5 g .200 -
E a. c 3 .160 - 2
.120 -
080 -
.040 -
.000
.040 i i i i . > >
.0000 .0036 .0071 .0107 .0143 .0179 .0214 .0250 Time, seconds O
5-9 Babcock & Wilcox
Figure 5-6. Transverse Section at Primary Piping OUTLET e INLET INLET A2 Al _y d
<M i x _ _
LOAD AT ' Z (VERTICAL) PLANE OF TOP OF RV SYMMETRY O l l l i O 5-10 Babcock & Wilcox
Figure 5-7. Support Skirt Detail Z LOAD h GLOBAL I COORDINATE SYSTEM H 45 l i / 14l X 15 21 22 / g 16 17 18 19 _ 20 g 5
@ g- g , G ,-
O @ / iN 2 O (i) 5 3 ELEMENT NUMBER N00E NUMBER l l O l 5-11 Babcock & Wilcox
O 6, UNIFORM LOAD ANALYSIS 6.1. Analytical Method The uniform load case assumed that the failure mechanism caused the falling head to hnpact the RV uniformly around the flange. This analysis used the same three-dimensional half-symmetry model of the RV. The drop height used was equal to the lift height for the point load case - 60 inches. Tha mass of the head was uniformly applied around the flange with an initial velocity to simulate impact. d.2. Results The response of the structure was monitored at the same locations as for the Predictable response was observed, and the maximum displace-
] point load case.
ments near the core flood nozzles are -0.027 inch horizontally inward (see Fig-ure 6-1) and -0.355 inch vertically downward (see Figure 6-2). The maximum displacements at the primary piping nozzle locations (see Figure 5-6) are as follows:
. Displacement, inches l Nozzle X Y Z l
Inlet A2 0.036 -0.023 -0.362 i j Outlet 0.008 -0.050 -0.358 Inlet A1 -0.040 -0.020 -0.362 i The stresses in the support skirt were assumed to be less than those in refer-ence 2, which were lower than the allowable limits set in the ASME Code." This assumption is based on the decreased drop height (from 120 to 60 inches) and the stiffness of the attached piping, which tends to decrease displacements, thereby decreasing the skirt stresses. The possibility of skirt buckling was dismissed based on the same arguments. 1 6-1 Babcock & Wilcox
Figure 6-1. Displacement Time History Near Cor: Flood Nozzle Radial Component, Uniform Load Case
.0480 -
.0400 -
.0320 -
.0240 -
E i 5e (
. 0160 - .
$ \ \
E s
- 5 .0080 - ! f N
I
.5
.0000 -
N
.0080 -
i
.0160 -
W
.0240 -
.0320
.0000 .0029 .0057 .0086 0114 .0143 .0171 .0200 Time, seconds O
6-2 Babcock & Wilcox
Figure 6-2. Displacement Time History Near Core Flood Nozzle Vertical Component, Uniform Load Case
.040 -
.000
.040
, .080 -
E 8
,- .120 -
5 B E .160 - 5 A E
.200 -
.240 -
.280 -
l l
.320 -
.360 ' ' ' ' ' ' '
.0000 .0029 .0057 .0086 .0114 .0143 .0171 .0200 Time, seconds O
6-3 Babcock & Wilcox
O
- 7. REACTOR COOLANT PIPING STRESSES s
The stresses developed in the reactor coolant piping as a resble of the heavy load displacements were analyzed for both the point load and uniform load cases. The displacements"3 '"" were used to determine the resultant moments along the piping runs."7 These moments were then compared to the allowable moments as calculated according to the 1977 ASME Code, Section III, paragraph NP-3652, equation 9, for Level D stress limits." The maximum resultant mo-ments were all lower than the allowable moments. See section 8 of this re-port for a summary cf these results. Based on these findings, the stresses developed in the reactor coolant piping duc to the heavy load drops are acceptable. The stresses developed in the E42-O $wd A ' 8 ( ,) core flood piping are to be analyzed by Florida Power Company or their desig 1 nees. l l l l ( I l l I I') s-I l l l l 7-1 Babcock &iMilcox 1 .-. ._ . -_ _ _ - . - - . .
O
- 8.
SUMMARY
AND CONCLUSIONS An analysis of heavy load drops on the reactor vessel has been performed ac-cording to the guidelines of NUREG-0612.1 The analysis considered the reactor vessel head falling obliquely to the plane of the vessel flange, causing point contact. It was assumed that the center of gravity of the RV head and attach-ments was directly above the point of contact. Also considered was the effect of the head falling without rotation from a level position and making uniform contact with the RV flange. Both analyses assumed a 60-inch lift height. The results from the two loading cases are summarized in Table 8-1. The results shown in Table 8-1 are conservative since for the most part they are obtained directly from analysis performed for the Oconee plant.s2 The Oconee RV stresses and nozzle deflections are bounding for Crystal River since the total weight that would be dropped on the Oconee reactor vessel is about 5000 lbs heavier than that for Crystal River (section 3.1). Also, the Crystal River rea;. tor coolant cold leg piping was found in reference 47 to be stiffer than that for Oconee' for vertical loads, which would further decrease RV stresses and deflections. The primary piping moments in Table 8-1 have been adjusted conservatively to reflect the Crystal River plant. This was done by applying the Oconee nozzle deflections to the stiffer Crystal River piping, resulting in upper bound loads in the pipes. l Based on results from the point and uniform load drop analyses, a 60-inch maxi-j mum lift of the RV closure head above the RV flange is recommended. l The drop of the upper plenum assembly is considered to be bounded by the drop l of the RV head on the RV flange considering equal drop heights. ! This conclusion is reached for two reasons. First, for non-uniform drops on the RV flange the impact of the plenum is less than the RV head due to its weight being slightly over one-third of the weight of the RV head. Second, a l O' V uniform drop of the plenum.on the reactor internals is considered to be an j event of low probability. It is unlikely that the plenum could sustain an ( 8-1 Babcock & Wilcox l - - - -
1 l exactly vertical drop into the vessel due to a failure in the lifting equip-O- ment, since the clearance between the keyways of the RV and upper plenum is such that any non-vertical drop would result in the plenum binding in the key-ways and thus the loading of the core support shield flange would be minimal. The drop of the upper plenum on the RV flange and core support flange is con-sidered to be bounded by the fuel handling accident for radiological release and the RV head drop on the RV flange for maintenance of. inventory. O l \ i l i L O 8-2 Babcock & VVilcox
' Analysis of Heavy Load Drops -- Summary at Results Table 8-1.
Point Uniform -Acceptance Analysis condition load load criteria RV head lift height, in. 60 60 NA General primary membrane stress, psi 44,375 (a) 49,000 Primary membrane plus bending stress, psi 44,493 (a) 73,500 Compressive load in support skirt, 10' lb 49.5 (a) 751 Nozzle deflections, in. Core flood nozzle Downward vertical 0.218 0.355 By FPC b#' Outward parallel to plane of symmetry 0.254 -0.027 By FPC AphI Outward perpend. to plane of symmetry 0 0 By FPC Reactor coolant cold leg A2 Downward vertical 0.193 0.362 (b) Outward parallel to plane of symmetry 0.220 0.036 (b) Outward perpend. to plane of symmetry 0.017 0.023 (b) Reactor coolant cold leg Al ('#) Downward vertical 0.067 0.362 (b) Outward parallel to plane of symmetry 0.181 -0.040 (b) Outward perpend. to plane of symmetry -0.008 0.020 (b) Reactor coolant hot leg Downward vertical 0.099 0.358 (b) Outward parallel to plane of symmetry 0.183 0.008 (b) Outward perpend. to plane of symmetry -0.120 0.050 (b) Resultant maximum moment in primary piping, ft-kips Reactor coolant cold leg A2 3614.2 4283.0 5841.4 Reactor coolant cold leg Al 1013.6 4467.7 5841.4 Reactor coolant hot leg 2210.9 6826.2 11876.0 (* Not calculated; assumed acceptable based on results frem reference 2. Acceptance criterion based on the stresses due to the resultant maximum moment through the piping runs, 4 x 8-3 Babcock & Wilcox
O APPENDIX Material Properties and Allowable Stresses O , A-1 Babcock & Wilcox
- 1. Bilinear Stress-Strain Data The nonlinear stress-strain curves used in the analysis for A-508 Class 2 and SA-516 Grade 70 are shown in Figures A-1 through A-3. These curves, provided by reference 2, are based on tests performed at B&W's Alliance Research Center in 1976 and 1979. As required by the ASME Code *, the curves are adjusted to correspond to the tabulated values of yield stress in the Code. ANSYS uses the energy of distortion (maximum octahedral shear stress) method in elastic-plastic analysis, which the Code accepts. This method requires the use of true stress-strain curves rather than the nominal stress-strain curves shown.
For that reason, the values obtained from the curves are adjusted to the val-ues of corresponding true stress-strain curves. 1.1. A-508 Class 2 (Reactor Vessel) From 1977 ASME Code : cy = 50,000 psi E = 29.9 x 108 psi From stress-strain curves: c = 0.108 o = 79,249 psi c = = = .001672 29 9 x 06 Stress, True stress, Strain, c e a(1+c) Yield 0.001672 50,000 50,084 Ultimate 0.108 79,249 87,808 1.2. SA-516 Grade 70 (Support Skirt) From 1977 ASME Code": cy = 38,000 psi E = 27.9 x 108 psi
*Section III, Appendix F, paragraphs F-1321. lb and F-1321.lc.
A-2 Babcock & Wilcox _ _ _ _ . . _ . .~ __ _ __.
from stress-strain curves: c = 0.158 @ -20F cu= 0.152 @ 125F e = 85,552 @ -20F o = 76,160 @ 125F Linearly interpolate to 70F: cu= 0.1543 @ 70F o = 79,722 @ 70F c = = '
= 0.001362 7 27 9 05 Stress, True stress, Strain, c a c(1+c)
Yield 0.001362 38,000 38,052 Ultimate 0.1543 79,722 92,023 0 1.3. SA-533 Grade B (Lower Head) From 1977 ASME Code": oy = 50,000 psi l E = 29.9 x 108 psi Since the lower head material does not yield in the analysis, no nonlinear material properties are calculated.
- 2. Code Allowable Stresses i
The following limits are imposed by the ASME Code, Appendix F: General primary membrane stress intensity (P,): 1.0 S, Local primary membrane stress intensity (P ): 1.5 S, Primary mcmbrane + primary bending stress intensity (P +P b ): 1.5 S, The S,value is calculated as the larger of
- 1. 0.7 S u O and
- 2. S + 31 (Su - S ),
y y Babcock & Wilcox _3
The following values of S and Sy are taken from Section III, Appendix I":
! S , psi S , psi 0.7 S u, S y+y(S-S),u y Material u y psi psi A-508 Class 2 80,000 50,000 56,000 60,000 SA-533 Grade B 80,000 50,000 56,000 60,000 SA-516 Grade 70 70,000 38,000 49,000 48,667 These values yield the following stress limits:
PL+Pb Pm (1.0 Sm) , PL (1.5 S.), (1.5 Sm). Material psi psi psi A-508 Class 2 60,000 90,000 90,000 l SA-533 Grade B 60,000 90,000 90,000 SA-516 Grade 70 49,000 73,500 73,500 i O i I
- O l
g_4 Babcock & Wilcox
O O O Figure A-1. Stress-Strain Curve for SA-508 Class 2 at 73F
! 80,000
- d uts 60,000 _ i ! d
'3
! .O v E
= 40,000 -
1 a S ! 5 l l t d = 53223 PSI w Ys 20,000 - duts = 79249 PSI I O ' I I I i 0 .05 .10 .15 .20 .25 Strain (in/in) I h i
?
O 1 e. w i .
1 O O O d Figure A-2. Stress-Strain Curve for SA-516 Crade 70 at 20F 4 j l i 80,000 - j 1 d uts i i i m 60,000 - w i 8 l w w
=
T O i * " d ys 40,000 . d ys = 52301 PSI i i i~ 20,000 - d uts = 85552 PSI l i i i,
. E O I ' ' ' '
l l o 0 .05 .10 .15 .20 .25 .30 a 1 X" Strain l e. l :6 4 = 1 a 4 O i 1 l l
Q O O Figure A-3. Stress-Strain Curve for SA-516 Grade 70 at 125F 80,000
] -
d uts 60,000 _
'; I 5
us 40,000 ,. 3 Ld ys d ys 49260 PSI
, duts = 76160 PSI i
9 20,000 - l 1 0 I I i i i . l ! - 0 .05 .10 , .15 .20 .25 .30 i i i Strain (in./in.) lmm _ 8 x h
~
w w o i
~
~
.. -- L
O REFERENCES 1 Control of Heavy Loads of Nuclear Power Plants, NUREG-0612 (1980). Babcock & Wilcox Calculation Data Sheet 32-1130504-00, " Heavy Loads Analy-sis - Reacter Vessel Head Drop, Midland Units 1 and 2," January 8, 1982. 3 C. J. DeSalvo and J. A. Swanson, ANSYS Engineering Analysis System User's , Manual, Vols I, II, Swanson Analysis Systems, Inc. , Houston, Pennsylvania.
" ASME Boiler and Pressure Vessel Code, 1977 Edition.
s R. W. Ganthner to R. L. Gill, Memorandum, " Inputs for RV Head Drop Analy-sis," Bebcock & Wilcox, January 27, 1982. T Babcock & Wilcox Calculation Transmittal Sheet 86-1130844-00, "RV Head Drop - Pipe Matrix," February 4,1982.
" Internals Handling E.. tension," Babcock & Wilcox Dwg No. 105737D11.
e
" Turnbuckle Pendant Assembly and Detail," Babcock & Wilcox Dwg No. 133303C2.
" Head and Internals Handling Fixture Assembly," Babcock & Wilcox Dwg No.
167221EO. 18 " Head and Internals Handling Fixture Details," Babcock & Wilcox Dwg No. 167222EO. 11 " Material List, Head, and Vessel." Babcock & Wilcox Dwg No. 149903E7. i, 12 Babcock & Wilcox l'*' '
"Specificatiott Drawing, Reactor Vessel Details, Sheet 1,"
Dwg No. 13618'1E3. 18 I ,
" Upper Shell Forging," Babcock & Wilcox Dwg No. 99319D1.
F." 1" " Arrangement, Reactor Vessel, Longitudinal Section," Babcock & Wilcox Dwg
't No. 149901E6.
l' " Lower Shell Forging," Babcock & Wilcox Dwg No. 104174D2. i B-1 Babcock & Wilcox
3 l' " Specification Drawing, Reactor Vessel Arrangement," Babcock & Wilcox Dvg No. 130180E3. 17 " Closure Head Assembly," Babcock & Wilcox Dwg No. 149919E4. is " Guide Studs and Seal Plug," Babcock & Wilcox Dwg No. 149925E2.
" " Upper Shell Assembly," Babcock & Wilcox Dwg No. 149904ES.
2o " Material List, Head c.nd Vessel," Babcock & Wilcox Dwg No. 152000E8. 21 " Closure Head Service Structure," Babcock & Wilcox Dwg No. 140991E3. 22 " Closure Head Service Structure Assembly, MK-329 Arrangement Plan View " Babcock & Wilcox Dwg No. 140998E2. 2s " Guide Studs and Seal Plug," Babcock & Wilcox Dwg No. 152010E2. 2" "Shell Assembly and Head Details," Babcock & Wilcox Dwg No. 151993E3. 2s " Upper Shell Assembly " Babcock & Wilcox Dwg No. 151901E11. 2s
" Upper Shell Assembly," Babcock & Wilcox Dwg No. 128737E6.
27 " Lower Head Forging " Babcock & Wilcox Dwg No. 99359D3. 2s " Upper Shell Forging," Babcock & Wilcox Dwg No. 99318D4. 2s " Upper Shell Forging," Babcock & Wilcox Dwg No. 99319D4. 3' " Seal Plate and Shield Plate Details," Babcock & Wilcox Dwg No. 167193E1. 31 " Arrangement, Seal Plate and Shield Plate," Babcock & Wilcox Dwg No. 167194EO. 32 " Vessel Head and Support Assembly and Detail " Babcock & Wilcox Dwg No. 128717E9. 33 "Shell Assembly and Head Details," Babcock & Wilcox Dwg No. 128705E9. 3" " Arrangement, Reactor Vessel, Longitudinal Section," Babcock & Wilcox Dwg No. 128702E14. ss " Material List, Head and Vessel," Babcock & Wilcox Dwg No. 128713E10. 35 " Upper Shell Forging," Babcock & Wilcox Dwg No. 99358D1. 37 " Vessel Flange Forging," Babcock & Wilcox Dwg No. 94929C4 38
" Upper Shell Assembly," Babcock & Wilcox Dwg No. 128704E9.
B-2 Babcock a Wilcox
3'
" Arrangement, Reactor Vessel Sections," Babcock & Wilcox Dwg No. 128703E17.
" Lower Head Forging," Babcock & Wilcox Dwg No. 99359D3.
"1
" Handling Fixture Sling, Material List " Babcock & Wilcox Dwg No. 78407BO.
42 " Head and Internals Handling Equipment " Equipment Specification 08-1005162-00, Babcock & Wilcox August 8, 1978.
"3 " Duke Heavy Loads Analysis - Nozzle Displacements," Calculation Transmit-tal Sheet 86-1131517-00, Babcock & Wilcox.
" Duke Heavy Load Analysis - Nozzle Displacements, Uniform Load Case," Cal-culation Transmittal Sheet 86-1131542-00, Babcock & Wilcox.
*s "RV Head Drop - Piping Loads Summary," Calculation Transmittal Sheet 85-1131550-00, Babcock & Wilcox, March 8, 1982.
Deleted.
"7
" Heavy Loads Analysis - Reactor Vessel Head Drop - Crystal River," Calcu-lation Data Sheet 32-1132113-01, Babcock & Wilcox, August 1983.
"' J. V. Vattamattam to J. A. Castanes, Memorandum, " Task 274, Rev. I and 2 W.A. No. 23," WPN 83-0870, Babcock & Wilcox #0096, August 23, 1983.
" Arrangement, Reactor Vessel Long. Section,'.' Babcock & Wilcox Dwg No.
126951E8, 620-0007. so
" Owner's Group Dimensional Comparison," Calculation Data Sheet 32-1106944-00, Babcock & Wilcox, February 19, 1980.
'l " Material List Head and Vessel." Babcock & Wilcox Dwg No. 135559E9, 620-0007-5H52.
s2 " Heavy Loads Analysis - Reactor Vessel Head Drop - Oconee Units 1, 2, and 3," Calculation Data Sheet 32-1132113-00, Babcock & Wilcox, March 22, 1982, O i B-3 Babcock & Wilcox
50TED r o/i2. 1983 I.J.LANZA vSubcocJt & Wilcox um, e , - eesseurass assuommy sais ow ro no.d Septamber 12, 1983
~
P.o.as 12eo . ESC-399 gg4 sos-12eo Mr,1. ylor Florida hr Corporation . P. O. Box 14042 St. Mtersburg, FL 33733 Attention: W . J. V Vattamattaa
Subject:
trystal River Unit 3 Master Services Agreement Effective Date: May 10,1978, B&W No. 582-7087 Task 274. Rev. 02 - Analysis of the Effect on the Reactor Vessel of RV Head Drop ..
References:
- 1) B&W Letter from J. A. Castanes to E. E. Renfro, " Task 274 Rev. 12 - Analysis of the Effect on the Reactor Vessel of RV seed Drop," FPC-83-237. ESC-271, dated June 15, 1983.
~2) TPC JdA; Contract NUC 10249AQ, WA No. 23
Dear Mr. -Taylor:
Attached is a brief evaluation of the Stone & Webster RV head drop scenario that results in an 1800 ft.-kips impact energy on the RV flange. B&W agreed to determine if the Consumers Power Company's point load head drop analysis could be used to evaluate the acceptability of the Stone & Webster postulated head drop. With this evaluation completed and acceptable results obtained, this transmittal completes Task 274, Rev. 02. A telecopy tyf the attachment has been forwarded to Mr. Tom.Lanza of Stone & Webster. . Very truly yo es, J. F. Walters, product Manager Owners Group Engineering Services cc: E. M. : Good E. M. Howard O J. A. Castanes R. J. Finnin D. W. Montgomery S
( . MORfM .
~
tone and Webster has postulated for Florida Power Corporation a RV head drop where-the AV-head end support structure is raised to the top of the,sh, allow end of the fuelensfer canal, the head drops, bouncing..off the canal wall onto the RV. The apper support structure strikes the RV at a single point and imparts an
. energy of 1800 ft.-kips to the itessel flange.
B&W was asked to evaluate this head drop scenario and determine if the CPCo Midland point load head drop analysis could be used to pass engineering judgement on the acceptability of the Stone and Webster postulated head drop. APPROACH Since the Midland plants contain bumpers around the upper vessel which limits lateral vessel motion under asymetric LOCA loads and the CR-3 vessel does not, the CR-3 RV can potentially displace acre in the lateral direction than the idland vessel. Additionally, the primary piping stiffnesses were not considered n the CPCo case. If including-the primary piping stiffnesses for CR-3 compensates for the lack of bumpers, the4tidland analysis can be.used as a basis for , evaluating the CR-31800 ft.-kips head drop case. Thus, comparing the.actualanergyabsorbed by the bumpers in the Midland analysis with the energy that could be absorbed by the CR-3 primary piping for the 1800 ft.-kips, a judgement could possibly be unde. Such a comparison revealed that the CR-3 primary system piping was able to provide greater support than the Midland bumpers, without overstenssing the RC piping. Considering the effect on the stresses in the RV and RV support skirt for the CR-31800 ft.-kips case, sufficient margin exists in the Midland stresses to accomodate the 8% increase in the CR-3 impact energy over the energy cpnsidered for the 5-foot Midland point load head drop analysis, without exceeding stress limits. . .
CONCLUSIONS Based on the Midland point load head dmp analysis, TRsference 1) some conputational effort and engineering judpent, the CR-31800 ft.-kip point load case may be considered to satisfy the requirements of Section 5.1 of NUREG-0612 (Reference 2). ' REFERENCES .
- 1. B&W D':ument 77-113hi20-DO, " Analysis of the Effect of Reactor Vessel Head Drop on the Reactor Vessel." December 1981.
- 2. NUREG-0612. "Cantrol of Heavy Loads at Nuclear Power Plants," 1980.
o S I e
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[ I ! l 4 ' APPENDIX G 5 I
- 1. Load Drop Analysis for RCCR-1 I
I 2. Load Drop Analysis for CWCR-1 l } 1 3. Load Drop Analysis for SFHT-7 [ ] : 1
- 4. Load Drop Analysis for RCCR-2 l- 5. Load Drop Analysis for FHCR-7 l
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STONE & WEBSTER ENGINEERING CCRPOR ATION CALCULATION SHEET ,, o ,, y, ,,y,,,,, ,,,,
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6"w/ TIT LEsrxWaas loAoDeco Aavets I /asca (~ Accumcrious . i.In%l veicdi4y o 0 enssie uaben dropped h seco z.'The miss(e eheikes +he, tocupt norrncd 4o +be surece.
- s. Any inbr enedake 4acc3eh ore. iojored.Pnmary *artte+
ickes Tuit impact. 4.hdo cruehnc3 o? the misele is used ed 'on i :n +he e cane
- s. trave.\
The load areaenay excep+be1.droe8nece.a+ any tophysica (n+eeference. is p G. \4 dradorces are. pre.sen* , they may be constdered. LiG+hech+ ie assumed +o be the maxthqm +ha+ t s phycictily possible, unless c4heruaise emad J based n a Cailurecs ,( I. The mos+ cri+ical cond:4ianonalp tdere ed. All o+ hec condi+to cri+M . E Anolysis is based en a bliinear elas+ic- p\cdic curve 4ho+ ne.presente, o true Wess /stcain retcktonship : io areigrcre , 's c4 exishh
\nlally +dbe un if4bg,.,
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. lb n passina ,+hudbe checkhd extsnng their ike etrecks % c.ri (ticcicads
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CALCULATlON SHEET , ,, , , c , , , , , ,, , , ,, , a g,, y,y,,,,, ,,,, m oet _ - 1J /2%.t'7 fFFC C-ol O foA E INDEPENDENT REVIE1EER /DATE IPPAEP RAER/ ,m s/ m e REVIEWE a m , sa. R / CMECat E RlLATE .
. %c-Av'es hM% W . w .s JR.T/ TITLE Q A CATEGORY / CODE CL ASS RvsiALNVR':5 bDCAbP AWAu<5E I /Ns4 Assumricas .
us . checked as a cola n wri% M/r C 22 S-Icuc+aJg leog% e / membecech are ic3nor ed ,asal \oads de4(e a Columas will -Ars+ be checkd igr.ccing lengh eL's
/ s. Passin columns u.hll caoin be checLcd u>&h lengR ehed if it is er rkcet
- m. (nfiol defledien+o be censdered uaie (cod dcgp deAlections.
u\\t inc\ade dead \oad c.nd c>ny perman:inMyb%:ns wed (cad u.ai\\ such as equt> ment, Earsguake cree.p , enc. , de s be negected. 3 k CCmbodico o E s-{ruchres may be prsse.Tk o resisi ihe lood d, cop., bouaevec 40 e mpl4y cmalysis cmly o ne. s4tuc-race. enay be onc yuJ l A MOre, boo AG, 6 6. Incn'n IOud k(CV'y bo C. % more c. rih \ toaIicC,she$".Ge/Abkbeindoep'/Soruted .Fet anotw e . um. . . - . . - n S!ruc4uro.l CMeed mppc4g concrehe, slabs dom nef be. ine.li4A cn de.veh.oig bareier ree;rane
' on verf l e, ses,smic FPC,lCRs response FSAR. is osamed Gec,Y +o onbe5.4.5',i dEeemgni iconi .Yha no induced accel siena oWer Won geavey ace considued occi 1
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% M e.rckio Q Ec i4 Qhs.umed ko re_(Aoiq- .congd_n{..during o trApck resishncs,ony increase h Ec due.+o dynams increase hter As assumed .ne3;opbae \ . . ._._. _
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- AaAwsts Ae presented in Re?.*l 3 +he eWechve rnass-p\cstic t%poc.t me+he crr +acop.t e+d noc+. is une.s used to in miseJes analysing 'dupped' 4he reponse. . corn cranes.
Tni m%cd evoluc4es an eC?ecMve moes 4er +he due.@on:R barrier An64recis +he import c.s a .
-bs+ic__ col \isbn between +ne missie and +he e@ec. tve, fiw e s, \ h e 61r d n e w of he -tocqst cd roaximum rt:.sponse is used 40 ance +be repiduca kan'ede_.
of 4he incqsk-m%sia cornbincTion mos s t4m , scag l _ne. impcct and si6ki veloc ,4,irnpoc inc5 a spri -bcck.ed 'is t9sdeled as +o dusste. og 3 gn3er masg Me,qs e spcim a biknear no a g unc.rien cT -me res(stanc.e-displac.tmeert properttes o-r ihe toncych J ,I For plashc.,couMcns v;Rh ei, odd echton Imgocks te. ( Tark.4 dpp\ccernents and spring . rces cce smad, dTream n dp na scopad and con cons tv:stv asn 4he be nea same.jecme masses s e.he end cm imMhs\e gnd cxst3
-t . r>etaretrcdn ener rs '
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STONE a wEssTER ENolNEERING CORPORATION CALCULATlON SHEET J.O./0.0. / CALCUL ATis m tg. '=c
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+ar =%loed.ne d 5n e:ms')
erc%n: Mcciog: es, enlssle gasses 4hrough the +arq9,$c.non obd cial behon: cie#cce;nent o, the f enkde icho vne +arof GeoWg: eleebn or udena\ 4com 4hc bra Cace n htgei Ecabbi)e&* becaus u.AU be +he gonicelhng localised condlenon6 ' d scabb:ng occurs ra beTere o pemorciton; c4eme on +be. bac.k side or p c g e t % c tt. Ehear o9 + spit % pse or noomokesT g in 44e ,p e, ds per+
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i TEST RLM FOR THEORTICAL APPROACH---COWARISON FOR Hate CALCULATION I wen m enenena n n en ume n a n n e n a n n o n-n om un u s u n m a n n a m on m o n n en nen n oon-n um e .um a no P 3 DATA CH HISSILE. BARRIER. Ate LOAD C0teINATION EQUATION meneseeenemannemoneenmanamene enamemm u umam enan u mann ee m am en e m mm mmmm mm m mm m mmmennen eem meeeeeeememem amenenaman nen BARRIER FORCE DISPLACEHENT RELATIONSHIP HIPS FEET T 3815.0 0.0100 3815.0 0.1000 3815.0 0.5000 , l) 0.0 HIPS EQUIVALENT STATIC FORCE em LOAD 1 O.0 MIPS EQUIVALENT CONSTANT DYHAHIC FORCE me LOAD 2 l l 0.0 HIP-SEC HISSILE IWLLSE RESISTED BY FORCE AT BARRIER SUPPORT PLUS BARRIER INERTIA DL5 TING HONErmM TRANSFER en LOAD 3 # ,) i 78.740 MIP-SEC HISSILE ItPULSE RESISTED OtA.Y BY BARRIER INERTIA DURING H0HENTIM TRANSFER en LOAD 4 24.4 FPS BARRIER INITIAL VELOCITY DUE TO LOAD 4
/
') BARRIER HISSILE HISSILE HEIGHT BARRIER BARRIER PLASTIC EFFEC. YIELD BARRIER PERICO l l EQUIVALENT HEIGHT HEIGHT LOAD 3 LOAD 4 FORCE DEFLECTION
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C BARRIER EQUIVALENT HISSILE HEIGHT HISSILE HEIGHT BARRIER PLASTIC EFFEC YIELD BARRIER BARRIER PERIOD . { HEIGHT LOAD 3 LOAD 4 FORCE DEFLECTION O HIPS HIPS HIF3 HIPS FT SEC - dl 31.420 0.000 174.000 3455.0 0.0024 0.0052 ** h a a m mm m m mm m mmmm m m m m mmm m mm m m m m mmm mmm m mmmm mmmmmm m m mmmm m m mm m m m m m m m m mm m m m m m m m m m m m m mm m m m m m m mm m mm mm m mmm m m m m m m m m men a m a n na ae RESULTS OF TIHE HISTORY ANALYSIS FOR HISSILE ItPACT HITH OTHER LOADS -l_L q y , m u u m m m m m mmm m m mm m m mm m m m mm m mmm m m m m mmmm m m mm mmm m mm mm m m m m m m m m m mm m m m m m m m m m m m m m m m m m mm mm mu s u m a m mu m m m mm m m m m mm m m m m m mAm mm u m l - i- j ~b ~ l V 1 2 3 4 5 6 7 8 9 TIHE HISTORY DURATIDH OF HISSILE FORCE FORCE AT TIME OF HAX BARRIER BARRIER HAXIHUh BARRIER HAXIHtM BARRIER hAXIHUM BARRIER FINAL BARRIER RESISTING HECHANISH b j Q NUteER LOAD 3 LOAD 3 SUPPORT DEFLECTIOH DEFLECTION DUCTILITY VELOCITY V SEC HIPS HIPS SEC FT FT/SEC @ p-J
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,e ! 3..M!m ..g s8.. a g ,= = e z K" ...r :.on 3..) 3. 3. e8..z 2..E - 25 444R 444r 5 R 444 g 444- d44R 444 5 - 5 E 444 !W dediEs 444 Rg - = "# 3- 0 5 i 0 W *E 4..* 4..: 4..g f..E j..* j. 5 4..g- 4..g 4.. g :E- 344E 344* g 344s g u44'd uddE uddl~- 3448~- Edds~ 344 a a d g ~ 3* g a g : I .... .... i . ....a 5( 5 s %g....- ...g ...g....-....go... .4444EdaddadaddedaddE4444s44ddedaddddaddddadd i i l' : C O C O O O O O O O O O 3 D J a6 - V ST-331.58tMI. VERS-00 LEVEL-00 71172 13.53.27 *\ -j e h -' PUte AT HAX HEIGHT RLM l' ..n o..u..u n n.o u n...-.n... .. ...- - o.... . .. . .......u n.n o - o.. ... . O O DATA ON HISSILEe BARRIER, Ate LOAD ColeIHATION EQUATIDH .... ...... ....e............... ..................................................................e....... O BARRIER FORCE DISPLACEHENT RELATIONSHIP HIPS FEET 747.0 0.0045 747.0 0.0450 747.0 10.0000 0.0 MIPS EQUIVALENT STATIC FORCE .. LOAO I 0.0 MIPS EQUIVALENT $0NSTANT DYHAHIC FORCE .. LOAD 2 O 0.0 MIP-SEC HISSILE It0MA.SE RESISTED BY FORCE AT BARRIER SUPPORT PLUS BARRIER IHERTIA DENIING MOHENTIM TRANSFER .. LOAO 3 St.a00 HIP-SEC HISSItE IseutSE RESISTEo Ora.Y af nARRIER IHERTIA OuRING H0HEHiuH TRANSrER .. LOAo = C l 27.4 FPS BARRIER INITIAL VELCCITY DUE TO LOAD 4 l sARRIER nARRIER O O nARRIER HISSILE EQUIVALENT. HEIGHT HISSILE HEIGHT sARRIER PLASTIC EFFEC. YIELD PERIOD I HEIGHT LOAD 3 LOAD 4 FORCE DEFLECTION j HIPS HIPS FT SEC C l 0 MIPS HIPS 4.140 0.000 00.000 767.0 0.0045 0.0044 l i,s O ......... ... .. .. ... ...... ............................,.............. ......n...... ...... RESULTS OF TIME HISTORY ANALYSTS FOR HISSILE INPACT HITH OTHER LOADS ...... .......... .. ................................o................................................... 10 - m .h s a 7 0 0 **- u ~<~. ~ h O ", O O I TIHE DURATION 3 HISSILE =. FORCE AT TIHE OF HAX HAXItAM HAXIttM HAXItAM FINAL BARRIER BARRIER BARRIER BARRIER BARRIER BARRIER RESISTING HECHANISH b HISTORY OF FOnCE HUBBER LOAO 3 LOAD 3 SUPPORT DEFLECTION DEFLECTION E XTILITY VELOCITY . 'h ,1 C Q ]N p HIPS HIPS SEC FT FT/SEC SEC 3=,.., s-SPECIAL .ARRIER SPRING ; D tf Q o
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.I p[6a 2 P t. p 4, g% v D ~3 0 v 'E ^ . 'se e n q / M ST-331.SBtMI. VERS-00 LEVEL-00 78.172 13.53.27 O O ,mHOTOR .- . . AT HAX HEIGHT RLA4 2 .......................................... ........................................e........................... O O DATA CH HISSILE. BARRIER. Ate LOAD CotmINATION EQUATION .e............................................................................................ ................ O O BARRIIR FORCE DISPLACEHENT RELATIONSHIP HIPS FEET O 9~ 747.0 0.0045 747.0 0.0450 747.0 10.0000 0.0 HIPS EQUIVALENT STATIC FORCE .. LOAD 1 0.0 HIPS EQUIVALENT CONSTAffT DTHAHIC FORCE .. LOAD 2 0.0 HIP-SEC HISSILE IPPLA.SE RESISTED BY FORCE AT BARRIER SUPPORT PLUS BARRIER INERTIA DL5 TING H0HENTIDt TRANSFER C .. LOAD 3 O 40.r00 HIP-SEC HISSILE IteutSE RESISTED OttY BT BARPIER INERTIA OuRIHo H0tiEnitAt TRutSrER .. LOAD 4 41.3 FPS BARRIER INITIAL VELOCITT DUE TO LOAD 4 HISSItE BARRIER BARRIER BARRIER C O BARRIER EQUIVALENT HISSILE HEIGHT HEIGHT PLASTIC EFFEC. YIELD PERIOD HEIGHT LOAD 3 LOAD 4 FORCE DEFLECTION HIPS HIPS rT SEC C O HIPS HIPS 4.140 0.000 40.000 747.0 0.0045 0.0064 C O ......................................................:........................................... ............. REStA.TS OF TIHE HISTORT ANALYSIS FOR HISSILE ItFACT HITH OTHER LOMS Q& e .p. % ................................................................................................................ . - . 1-8 9 O d I TIHE 2 DURATION 3 HISSILE 4 5 FORCE AT TIttE OF HAX 4 HAXIltki 7 HAXIIIUH HAXIHUH FINAL BARRIER \ k,g7 \j BARRIER BARRIER BARRIER BARRIER BARRIER RESISTItG HECHANI$H ,i HISTORY OF FORCE 00 NUtBER LOA 0 3 LOAD 3 SUPPORT DEFLECTION DEFLECTIOtt DUCTILITY VELOCITY * \ U 0 HIPS Occ FT FT/SEC (g @N {. 'j ~ SEC HIPS ( ?O p V 0 0.0 0.0 747.0 0.078484 1.4193 359.84 41.30 SPECIAL BARRIER SPRItG hC 1 w._ c- , P w .@ b % . bl V W e va 4, tfl I l l go ' G C L G L- O O O .O O O O 3 O J 4 h ce k P.h 4 *hto V:pc. / c.RS \&2.%.G C.-O \ , che.p.7,& A.p /n/b 5 R.W- . h-A G4,EL g = . . SA'f'$r}3th]!? m
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m = g = .5le -.3 e - E' I g gg a l 'e a s e ~ e ae s O Q 1 ST-331,5869fI. VERS-00 LEVEL-0010.172 13.53.27 . O C (Putt' AT ALLOM HEIGHT RtM 4 seeeeeeeeeeen.ne... ..een.nemes..meseeen.ameenenses.o..e..em.a.neen.....eu..meese....senes. wen.ne.e..me.neeees.e O O DATA DN HISSILE. BARRIER. Ate LOAD CateINATION EQUATION
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O BARRIER FORCE DISPLACEHENT RELATIONSHIP HIPS FEET 254.0 0.0028 254.0 0.0280 254.0 10.0000 0.0 HIPS EQUIVALENT STATIC FORCE .. LOAD 1 0.0 HIPS EQUIVALENT CONSTANT DYHAHIC FORCE .e LOAO 2 0.0 MIP-SEC HISSILE IFPLA.SE RESISTED BY FORCE AT BARRIER SUPPORT PLUS BARRIER It'ERTIA DtstING H0HENTIM TRANSFER s. LOAD 3 4.183 HIP-SEC HISSILE IteutSE RESISTEo Ota.y By BARRIER IHERTIA ouRIHG H0HEHrim TRaHSrER .e LOAo a C O 2.2 FPS BARRIER INITIAL VELOCITY DUE TO LOAD 4 BARRIER BARRIER C-C BARRIER EQUIVALENT HISSILE HEIGHT HISSILE HEIGHT BARRIER PLASTIC EFFEC. VIELO PERIOD HEIGHT LOAD 3 LOAD 4 FORCE DEFLECTIDH HIPS HIPS FT SEC C Q HIPS HIPS 1.350 0.000 90.000 254.0 0.0028 0.0043 C o ...e................es..e....e.e......... ......e.............................e... ..eeen. ........... .. ...ee RESULTS OF TIHE HISTORY AHALYSIS FOR HISSILE IHPACT HITH OTHER LOADS ... ........e............m............ ........................................................ ............... 4 5 4 7 8 0 "(.1 t O O 1 TIHE 2 DURATIM4 3 HISSILE FORCE AT TIME OF HAX HAXItRM HAXIHUH HAXIHUH FINAL BARRIER g)1 () HISTORY OF FORCE BARRIER BARRIER BARRIER BARRIER BARRIER RESISTING HECHANISH l *g* p .. , LOAD 3 SUPPORT DEFLECTION DEFLECTION DUCTILITY VELOCITY O, HutefR LOAD 3 < -aD D v e HIPS HIPS SEC FT FT/SEC ' SEC ~~ 0.0 2s4.0 0.029197 0.0277 9.90 2.18 SPECIAL BARRIER SPRING $ ' -i >1 .o*E O O U 0 0.0 +- -0) D{ i .h u gu 6m x'* =6 U P SI g W" .ee.. O Z ST-331.EP4941, VERS-00 LEVEL-00 78.172 13.53.27 r O ,e HOTOR AT ALLOH HEIGHT RUH 5 .............Hud................................................................s..............................e DATA CH HISSILE, BARRIER, AtW1 LOAD CateINATIDH EQUAT10tt O O O O BARRIER FORCE DISPLACEMENT RELATIONSHIP HIPS FEET D O 256.6 0.0028 256.0 0.0280 254.3 10.0000 0.0 MIPS EQL'IVALENT STATIC FOR(.E .. LOAD 1 0.0 MIPS EQUIVALEHT CONSTANT DYHf. HIC FORCE .. LOAD 2 0.0 HIP-SEC HISSILE IFFLASE RESISTED BT FORCE AT BARRIER SUPPORT PLUS BARRIER IHERTIA DL5 TING H0HENTIM TRANSFER ..C- LOAD 3 O ..zIt HIP-SEC HISSILE IirutSE RESISTED OrtY BT BARRIER IHERTIA DuRIts 90HEHTuH TRANSrER .. toad = 3.2 FPS BARRIER INITIAL VELOCITY DUE TO LOAD 4 HISSItE HISSItE BARRIER BARRIER BARRIER C O BARRIER EQUIVALENT HEIGHT HEIGHT PL ASTIC EFFEC. TIELD PERIDO HEIGHT LOAD 3 LOAD 4 FORCE DEFLECTION HIPS HIr5 rT SEC C O HIPS HIPS 1.350 0.000 40.e00 254.0 0.0028 0.0043 C O ................................................................................................................ REStA.TS OF TIHE HISTORY AHALYSIS FOR HISSILE DFACT HITH OTHER LOADS C. O z.. 4.................ww....................................................................................... 5 4 2 8 , O O I TIHE 2 DURATION 3 HISSILE FORCE AT TIHE OF HAX HAXIHUH HAXIHUH HAKIHUH FINAL BARRIER N 3g RESISTItG HECHANISH "f4 g pM BARRIER BARRIER BARRIER BARRIER BARRIER HISTORY OF FORCE HUHBER LOAD 3 LOAD 3 SLPPORT DEFLECTION DEFLECTION DUCTILITY VELOCITY - - Dyo V, .. . O SLC HIPS HIPS SEC FT FT/SEC l .N -g_ ) bO C 254.0 0.01491s 0.0280 ( 9.99 3.22 SPECIAL BARRIER SPRING ( e Q U 0 0.0 0.0 f #$e .
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s 4 J> L - t% 5 w =a p.. <. ~ I S O [ ,\ ST-331,$Bitt!. VERS-00 LEVEL-00 78.172 .13.53.27 ._ O I O ( LOG AT ALLOM HEIGHT RUN 4 ... ...............................................................................r............................ O DATA Ott HISSILE, BARRIER. APO LOAD COFetHATION EQUATION O D.. O BARRIER FDRCE DISPLACOlENT RELATI0ttSHIP HIPS FEET c . O I 2E5.0 0.0028 256.0 0.0280 256.0 10.0000 C O 0.0 MIPS EQUIVALENT STATIC FORCE .u LOAD 1 0.0 HIPS EQUIVALENT CottSTArn DYHANIC FORCE .. LOAD 2 0.0 HIP-SEC HISSILE ItFtLSE RESISTED BY FORCE AT BARRIER SUPPORT PLUS BARRIER IHERTIA DURING Hul.ENTitt TRANSFER b em LOAD O 2.050 HIP-SEC HISSILE ItPULSE RESISTED OtLY BY BARRIEk INERTIA DURING H0HENTUH TRAMSFER .. LOAD 4 4.4 FPS BARRIER INITIAL VELOCITY DUE TO LOAD 4 HISSItE HISSItE BARRIER BARRIER BARRIER C O BARRIER EQ'JIVALENT HEIGHT HEIGHT PLASTIC EFFEC. YIELD PERIOD HEIGHT LOAD 3 LOAD 4 FORCE DEFLECT 10tt HIPS HIPS FT SEC C O HIPS HIPS 1.350 0.000 8.700 254.0 0.0028 0.0043 C O ................................................................................................................ RESULTS OF TIME HISTORY ANALYSIS FOR HISSILE IIPACT HITH OTHER LOADS L O .rns.....................g...........................................>.......................................... 4 5 6 7 8 , O O I 2 HISSILE 3 FORCE AT TIHE OF HAX HAXIHUH HAXIHUH HAXIllUH FINAL BARRIER g3 y D . ,N y TIME DURATION BARRIER BARRIER BARRIER RESISTING HECHAHISH j HISTORY OF FORCE BARRIER BARRIER 4'g VD ~Q yq _s LOAD 3 LOAD 3 SUPPORT DEFLECTION DEFLECTI0tt DUCTILITY VELOCITY 'IUHBER e . fp - "V SEC HIPS HIPS SEC FT FT/SFC g I ~~ ,
- i. N 4.57 SPECIAL BARRIER SPRING b v 9.92
;* @C D, h -11 g 0.0 256.0 0.006219 0.0278 V O 0.0 N _./ . O 1> a v N q 4; - ,J} ('~b u .4 L 0 . M Oh c r0 N3 . _ . _ _ d aa - D D \ ( s ST-331.SBFilIeVERS-00 LEVEL-00 78.172 13.53.F7 r O ALTERHATE PUFF Allut. HEIGHT RUN 1, O DATA OH HISSILE, BARRIER. Ate LOAD COBINATION EQUATION O ................................................................................................................ r O BARRIER FORCE DISPLACEHENT RELATIONSHIP HIPS FEET b O 767.0 0.0015 767.0 0.0145 747.0 10.0000 D 0.0 MIPS EQUIVALENT STATIC FORCE .. LOAD 1 0.0 HIPS EQUIVALENT CONSTANT DYHAHi. FORCE .. LOAD 2 0.0 MIP-SEC HTSSILE IFFtLSE RESISTED BY FORCE AT BARRIER SUPPoaf PLUS BARRIER IHERTIA DLRIHG H0HEHTIAI TRANSFER p. LOAD 3 f C 7.370 HIP-SEC HISSILE ItFLA.SE RESISTED OtLY BY Bt.RRIER INERTIA DURItG H0HENTW TRANSFER .. LOAD 4 2.6 FPS BARRIER INITIAL VELOCITY DUC TO LOAD 4 r' HISSILE HISSItz BARRIER BARRIER BARRIER O BARRIER EQUIVALENT HEIGHT HEIGHT PLASTIC EFFEC. YIELD PERIOD j DEFLECTION HEIGHT LOAD 3 LOAD 4 FORCE HIPS HIPS #CIFS FT SEC , O HIPS 2.510 0.000 90.000 767.0 0.0015 0.0024 C O .................................................................................................=.............. RESULTS OF TIME HISTORY ANALYSIS FOR HISSILE IHPACT HITH OTHER LOADS C, O .................................................................................<.............................. 4 5 6 7 8 9 C O I 2 3 HAXIHUH FINAL BARRIER TIHE DURATIDH HISSILE FORCE AT TIHE OF HAX BARRIER BARRIER HAXIHUH BARRIER HAXIHUH BARRIER BARRIER RESISTING HECHANISH g f y 11 HISTORY OF FORCE ' p ys ~17 C HUF6ER LOAD 3 LOAD 3 SUPPORT DLFLECTION DEFLECTION DUCTILITY VELOCITY FT/SEC e 08g O ) V SEC HIPS HIPS SEC FT R{V ~ F U 0 0.0 0.0 767.0 0.009899 0.0131 9.02 2.57 SPECIAL BARRIER SPRItG N & .j G d %- 7- 6 ,m t%e to c . % -e,- Fi 00 2 66 G" \ 3 . O j ST-331.58094I. VERS-00 LEVEL-00 70.172 13.53.27 O s O - ALTERNATE HOTOR ALLOM HEIGHT RLM e ... . .... m . u...... ..................................... ............... .................... ... . O O DATA CH HISSILE. BARRIER. Ate LOAO COHBINATION EWATION . C O SARRIER FORCE DISPLACEHENT RELATIONSHIP HIPS FEET 747.0 0.0015 ui.0 0.0145 747.0 10.0000 0.0 HIPS EQUIVALENT STATIC FORCE .. LOAD 1 0.0 MIPS EQUIVALENT CONSTANT OYHAHIC FORCE .. LOAD 2 LOAD 3 j 0.0 HIP-SEC HISSILE IWULSE RESISTED BY FORCE AT BARRIER SUPPORT PLUS BARRIER IHERTIA DL5tIHS t0Ao . HDHENTW TRANSFER C l O ..,z, HIP-SEC HISSItE IW uLSc RESISTEo Ote.Y er sARRIER IHERTIA DURIHo H0HEHvim TRANSrER 3.7 FPS BARRIER INITIAL VELOCITY DOE TO LOAO 4 HISSItE sARRIER eARRIER .ARRIER C O sARRIER EQUIVALENT HISSILE HEIGHT HEIGHT PLASTIC EFFEC. YIELD PERIOD HEIGHT LOAD 3 LOAD 4 FORCE DEFLECTION hie 5 HIPS rT Sec C O HIPS HIPS 2.510 0.000 40.000 747.0 0.0015 0.0024 C O ... ....... ...... .. ....... ................................................ ....... .... RESULTS OF TIHE HISTORY ANALYSIS FOR HISSILE I W ACT HITN OTHER LOADS C O .......... ......... ....................................... ...................................... ...... O 1 2 3 4 5 6 7 HAXIIAJH 8 HAXIIAM 9 F MAL BARRIER p- .$m- 'y py HISSILE FORCE AT TIHE OF HAX HAXIHUH TIltE OLEIATION FORCE BARRIER BARRIER . BARRIER BARRIER BARRIER RESISTING HECHANISH ; -1 p HISTORY OF U tAhBER LOAD 3 LOAD 3 SUPPORT DEFLECTION DEFLECTION DUCTILITY VELOCITY , le s [s t. HIPS HIPS SEC FT FT/SEC . SEC r . P. o 0 0.0 0.0 u?.0 0.00ure 0.01:5 ,
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~. v0 0 $ JM aa O D { A \ 8 , ST-331.SB60t!. VERS-00 LEVEL-00 78.17E 13.53.27 n RHATE LOG ALLOH HEIGHT RUH 9 Os O DATA OH HISSILE. BARRIER. Ate LOAD C0tBINATION EQUATION ...............................................................r........................................ C I O BARRIER FORCE DISPLACEHENT RELATIONSHIP l HIPS FEET C O 767.0 0.001", 767.0 0.0145 767.0 10.0000 0.0 MIPS EQUIVALENT STATIC FORCE .. LOAD 1 0.0 HIPS EQUIVALENT CONSTANT DYHAHIC FORCE .. LOAD 2 m 0.0 MIP-SEC HISSILE ITtLSE RESISTED BY FORCE AT BARRIER SUPPORT PLUS BARRIER IHEttTIA DUEING N0HENTUH TRANSFER ( _ .. LOA O 2.460 HIP-SEC HISSILE IleutSE RESISTED OttY BY BARRIER INERTIA ouRIHo iniENTim TRAnSrER .. TOAD . 7.6 FPS BARRIER INITIAL VELOCIT / DUE TO LOAD 4 n HISSILE BARRIER PARRIER BARRIER C BARRIER EQUIVALENT HISSILE HEIGHT HEIGHT PLASTIC EFFEC. YIELD PERIOD HEIGHT LOAD 3 LOAD 9 FORCE DEFLECTION HIPS HIPS HIPS FT SEC 3 HIPS 2.S10 0.000 8.700 147.0 0.0015 0.0024 v 0, RESULTS OF TIHE HISTORY ANALYSIS FOR HISSILE ItPACT HITH OTliER LOADS b_ O ......................................s........................................z............m................... H =F -P 5 A O 4 5 6 7 8 9 4 _t O I TIHE 2 DURATIDH 3 HISSILE FORCE AT TIHE OF HAX HAXIIIUH HAXIltUH HAXIDE.H FIrlAL BARRIER p --g b RESISTING tlECHANISH , HISTORY OF FORCE BARRIER BARRIER BARRIER BARRIER BARRIER MP HUNDER LOAD 3 LOAD 3 SUPPORT DEFLECTION DEFLECTION DUCTILITY VELOCITY t1 O O SEC HIPS HIPS SEC FT FT/SEC lh- ]-D fi- % *Ng 747.0 0.003540 0.0141 9.69 $ 7.44 SPECIAL BARRIER SPRIT _, U 0 0.0 0.0 h _ s 0 % ( 7 W % p ellr o , v' if $ t ---- I CALCULATION SHEET R E V15103 J.O./0.0./ CALCpLATlYJ LJ. PAGE em (M.05.I'7/ FPG COI O S'luditCT/ TITLE / & -/ b3 $$ ~ $lSM $ h $l$
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O LOAD DROP S 1 ANALYSIS FOR SFHT7-OH HALL
.............. ............... . ............................................................ ..... C DATA CH HISSILE, BARRIER, Ate LOAD CotBINATIOH EQUATION O
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O 5.470 HIP.SEC HISSILE ItPtR.SE RESISTED OtLY BY BARRIER INERTIA CURING H0HEHTtM TRANSFER .. LOAD 4 21.9 FPS BARRIEW IHITIAL VELOCITY DUE T'J LOAD 4 BARRIER BARRIER C O BARRIER EQUIVALENT HISSILE HEIGHT HISSILE HEIGHT BARRIER PLASTIC EFFEC. YIELD PERICO HEIGHT LOAD 3 LOAD 4 FORCE DETLECTIO*l HIPS FT SEC C
C HIPS HIPS HIPS 4.425 D.000 3.900 3855.0 0.0I05 0.0038 O ....... ......... .... ........... .......................................................................
RESLA.TS OF TIHE HISTORY ANALYSIS FOR HISSILE ItfACT HITH OTHER LOADS O ......................................... .....................................................................
5 4 7 8 9 O 1 2 3 HISSILE 4
FORCE AT TIME OF HAX HAXIHUM HAXIHUM HAXIIRM FINAL BARRIER TIHE OURATION BARRIER BARRIER RESISTING HECHAHISH OF FORCE BARRIER BARRIER BARRIER HISTORY NUtBER LOAD 3 LOAD 3 SUPPORT DEFLECTION DEFLECTION DUCTILITY VELOCITY v U HIPS HIPS SEC FT FT/SEC SEC 3855.0 0.001720 0.0216 2.04 *1.93 SPECIAL BARRIER SPRING C O O 0.0 0.0 U ATTACHMENT Io CALC. NO. c-ol
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O LOAD DROP 0 2 ANALYSIS FOR SFHTT--OH SLAS HITHOUT HATER ORAG ...
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' O.0 HIPS EQUIVALENT CONSTANT DYHAHIC FORCE .. LOAD 2 0.0 HIP-SEC HISSILE It0'tA.SE RESISTED BY FORCE AT BARRIER StPPORT PLUS BARRIER INERTIA Ot5 TING H0HEHTut TRANSFER .. LOAD 3 C-O 4.700 HIP-SEC HISSILE ItFULSE RESISTED ONLY BY BARRIER IHERTIA DURING HDHENTIM TRANSFER .. LOAD 4 12.1 FPS BARRIER IHITIAL VELOCITY OUE TO LOAD =
t HISSItE .ARRIER eARRIcR sARRIER C l O nARRIER EQUIVALENT HEIGHT HISSItt HEIGHT PLASTIC EFFEC. YIELD PERIOD HEIGHT LOAD 3 LOAD 4 FORCE DEFLECTION HIPS HIPS HIPS FT SEC O HIPS 14.137 0.000 3.,00 4324.0 0.0015 0.0025 O mm . m ..... . m . m . . m . m m .. m .. m . m .. m . m ... m .. - . m m m .. . m RESULTS OF TIME HISTORY AHALYSIS FOR HISSILE It9ACY HITH OTHER LOADS O . m o. m . m .. m . m m ... m ...... m .. m .................... - .... m .. m ......... m . ....
O 1 2 3 = 5 4 i e , O HISSILE FORCE AT TIME OF HAX HAXIHUH HAXIEUH HAXIHUM FINAL BARRIER TIME DURATION FORCE BARRIER BARRIER BARRIER BARRIER BARRIER RESISTING HECHANISH HISTORY OF IAAeER LOAD 3 LOAD 3 SUPPORT DEFLECTIDH DEFLECTION DUCTILITY VELOCITY G, O SEC HIPS HIPS SEC FT FT/SEC U 0 0.0 0.0 4324.0 0.0014to 0.0104 4.91 It.Io SPECIAL BARRIER SPRING C U
%.' 2D. W"I* &//- ' ATTACHMENT fo T hv. 2/ CALC. No. c.or M! de PAGE 3 0Fy [
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O LOAD DROP O 2 ANALYSIE FOR SFHT7-OH SLAB HITH HATER DRAG neeeeeeeeeeeeeeeeeeeeeemmemammeenesseeeeeeeeeeeeemameneeeeeeeeeeeeeeemeneseeeeeeeeeeeeeeeeeeeeeenmoosenesee O
DATA CH HISSILE, BARRIERe Ate LOAD C0teINATION EQUATION C
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! 10.e7 SPECIAL sARRIER SPRING 0.0 4324.0 0.001400 0.00as 4.04 _
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1 4 l O 12345420,012345420,2012345470,30123454it,012345470,50123454it,4012345sie,7012345470,0 0 r l LOAD OROP ANALYSIS FOR RCCR2--CASE 014 54.4 FT & 0.3 HIPS 1
- 0. t 1.0 3455. 0.00244 3455. 0.0244 3455. 9.2 31.42 3 i
! O 0. 0.0 8. 0.0 4 D 0.0 0.0 0.0 0.57 0.30 5
< LOAD DROP ANALYSIS FOR RCCR2--CASE 02 4 33.4 FT & 0.3 HIPS 4 .
'l 0. 7 O 1.0 1,7.5 0.04244 1,7.5 0.424 1,7.5 1.0 1.e4 a O
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! 0.0 0.0 0.0 0.431 0.3C 10 1 LOAD DROP ANALYSIS FOR RCCat--CASE 01 4 56.4 FT & 2.75 MIPS 11 O 0. It C
, 1.0 3455. 0.00244 3455. 0.0244 3455. 0.t 31.42 13 4 0. 0.0 0. 0.0 14 O L6AD DROP EYSr5 FYRCCat E St a 5iS FT& t.75 MIPS Il17 C 0.
1.0 1,7.5 0.04244 1,7.5 0.424 197.5 1.0 1.,4 18
- 0. 0.0 8. 0.0 19
- O 0.0 0.0 0.0 3.94 2.750 to C
- LOAD DROP ANALYSIS FOR RCCR2--CASE 014 58.4 FT & 5.00 MIPS 21
- 0. 22 1.0 3455. 0.00244 3455. 0.0244 3455. 0.2 31.42 23 i O 0. 0.0 8. 0
,.0 24 C-
- 0.0 0.0 0.0 .48 5.00 25 j LOAD DROP ANALYSIS FOR RCCRE-CASE St & 33.4 FT & 5.00 MIPS 24 O. 27 i O 1.0 191.5 0.04
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ST-t:1.SBtNI.VE;S-00 LEVEL-00 73.172 13.53.27 O LOAD DROP ANALYSIS FOR RCCR2--CASE 01 4 50.4 FT & 0.3 HIPS
, m..... . .m m m m......-.m m.m m..m.......m......m.m.... .m m . .
O DATA OH HISSILE, BARRIER, Are LOAD C0FeINAT10H EQUATIDH r O P
' BARRIER FORCE DISPLACEMENT RELATIONSHIP 4
I HIPS FEET
- O C l 3455.0 0.0024 3455.0 0.0244 3455.0 0.2000 i
O C l 0.0 HIPS EQUIVALENT STATIC FORCE .. LOAD 1 j 0.0 HIPS EQUIVALENT CONSTANT DYHAHIC FORCE .. LOAD 2 0.0 HIP-SEC HISSILE IWLA.SE RESISTED BY FORCE AT BARRIER SUPPORT PLUS BARRIER INERTIA DL5 TING MONENTIM TRANSFER .. LOAD 3
, O 0.5T0 HIP-SEC HISSItE Im utSE RESISTED DetY By BARRIER INERTIA DuRING H0HEHTim TRANSFER .. TOAD 4 C l
0.4 FPS BARRIER INITIAL VELOCITY DUE TO LOAD 4 O BARRIER HISSrtE HISSILE BARRIER BARRIER PLASTIC EFFEC. YIELD BARRIER PERIOD C-EQUIVALENT HEIGHT HEIGHT HEIGHT LOAD 3 LOAD 4 FORCE DEFLECTION i
O HIPS HIPS HIPS HIPS rT SEC
; 31.420 0.000 0.300 3455.0 0.0024 0.0052 i o ......... ... ................. .................... ....................... - ....... ..............
RESLA.TS OF TIHE HISTORY ANALYSIS FOR HISSat.E Its'ACT HITH OTHER LOADS ..
O m m m ..m . . . .. .. .. m m .. .. . .-. .m . m . . . . . . . m m. . .. . . . .m . . - -
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TIHE DURATIDH HISSILE FORCE AT TIHE OF HAX HAXIHUH HAXIttJH HAXIHlh FIHAL BARRIER t HISTORY OF FORCE BARRIER BARRIER BARRIER BARRIER BARRIER RESISTING HECHANISH HUFBER LOAD 3 LOAD 3 $UPPORT DEFLECTION DEFLECTIDH DUCTILITY VELOCITY SEC HIPS HIPS SEC FT FT/SEC l
i i O O 0.0 0.0 489.7 0.001300 0.0005 0.20 0.54 SPECIAL BARRIER SPRIHG C U 9e 2'D.G/d W 83 ATTACHMENT O
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j 0 V ST-331.SBeeft.VE~S-00 LEVEL-00 73.172 13.53.27 O LOAD DROP AnAtvSIS r0R RCCR2--CASE Or a 33. rT a 0.3 HIPS r u.n.... noonn n .nunun.un..n.n.n.-nnuno..... .......unn.n===ne... a O DATA ON HISSILE. BARRIER. Ate LOAD C0tEINATIO't EQUATIDH I
. ..... ........ s ............. ............. .....L. m x...............u... ......... .........s..
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o BARRIER FORCE DISPLACEHENT RELATIONSHIP e
HIPS FEET g 197.5 0.0424 191.5 0.4240 197.5 1.0000 c l.
0.0 MIPS EQUIVALENT STATIC FORCE .. LOAD 1 j
0.0 HIPS EQUIVALENT CONSTANT DYNAHIC FORCE .. LOAD E O.0 MIP-SEC HISSILE It0'tm.SE RESISTED BY FORCE AT BARRIER SUPPORT PLUS BARRIER INERTIA OtstING HOHENTLSt TRANSFER .. LOAD 3 0 0.=31 HIP-SEC HISSILE ItetA.SE RESISTED Ota.Y By BARRIER INERTIA OuRIHs HoHENItti TRAnSrER .. LOAD
- C
< 4.1 FPS BARRIER INITIAL VELOCITY DUE TO LOAD 4 4 O BARRIER HISSILE MISSILE Bt.RRIER BARRIER BARRIER O EQUIVALENT HEIGHT HEIGHT PLASTIC EFFEC. YIELD PERIOD l HEIGHT LOAD 3 LOAD 4 FORCE DEFLECTION O HIPS HIPS HIPS HIPS rT SEC C I
1.960 0.000 0.300 197.5 0.0424 0.0227 0 .............. .. ......... ................................................... .....................
1
! REStA.TS OF TIHE HISTORY ANALYSIS FOR HISSILE ItFACT HITH OTHER LOADS O (
...........n...................................................on..
O 1 2 3 4 5 6 7 8 9 C
- TIHE DURATIDH HISSILE FORCE AT TIHE OF HAX HAXIHUM HAXIHUM MAXItttt FINAL BARRIER j HISTORY OF FORCE BARRIER BARRIER BARRIER BARRIER BARRIER RESISTING HECHANISH tRABER LOAD 3 LOAD 3 SUPPORT DEFLECTI0tt DEFLECTION DUCTILITY VELOCITY
! O b i SEC HTPS HIPS SEC FT FT/SEC I
Q 0 0.0 0.0 111.2 0.006100 0.0239 0.54 4.14 SPECIAL BARRIER SPRIHS U i
l C B p' 2 b Z.//.- ATTACHMENT 7 v l 7 -
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- O LOAD oROP AHr.:.vSIS r0R RCCR2--CASE 01 a se.. FT A 2.75 MIPS r messesseeeeeeeeeeeeeeeeeeeeeeeeeen eseeneson ne s eeseen meem seneses sesse s s e s sm e m enemmeneem sem assemememmeneenesessmen O oATA OH HISSItt, nARRIER. Are t0Ao C0teINATION EQUATION r eneseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeenommeneenmeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeneme l O O i BARRIER G)RCE DISPLACEMENT RELATION 5 HIP j HIPS FEET O C I
3455.0 0.0024
>j 3455.0 0.0244 it 3455.0 0.2000 i O C
- 9.0 MIPS EQUIVALENT STATIC FORCE se LOAD 1 j 0.0 HIPS EQUIVALENT CONSTANT OYHAHIC FORCE se LOAD 2 l 0.0 MIP-SEC HISSILE IWELSE RESISTED BY FORCE AT BARRIER SUPPORT PLUS BARRIER INERTIA DL5tIHB HOMEHitM TRANSFER em LOAD 3 I
O 5.210 HIP-SEC HISSILE IteutSE RESISTEo Ote.y av sARRIER INERTIA OuRIHs H0HENTim TRANSFER en t0A0 .
4.9 FPS BARRIER IMITIAL VELOCITY DUE TO LOAD 4 C.
O sARRIER HISSItt HISSILE sARRIER sARRIER sARRIER C i ! EQUIVALENT HEIGHT HEIGHT PLASTIC EFFEC. VIELD PERIOD i
l HEIGHT LOAD 3 LOAD 4 FORCE DEFLECTION I
O HIPS HIPS HIPS HIPS FT SEC C
!l 31.420 0.000 2.750 3455.0 0.0024 0.0052 i ,
O -enemenemm enemmemseen- m een enewes o me nemm e m-mes m o m e mm e m meenseen-messesee m se C
! RESLLTS OF TINI HISTORY ANALYSIS FOR HISSILE ItPACT HITH OTHER LOADS l O m m -wemeewoesene.ame- - mem mem - momem mmm meneem mem - memom m m e C
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1 Q 1 2 3 4 5 6 7 8 9 L TIHE DURATION HISSILE FORCE AT TIME OF HAX HAXIHUM HAXIHUH HAXIHUH FINAL BARRIER I HISTORY OF FORCE BARRIER BARRIER BARRIER BARRIER BARRIER RESISTINS HECHANISH
, taAeER LOAD 3 LOAD 3 SUPPORT DEFLECTION DEFLECTION DUCTILITY VELOCITY 4 O G
_ SEC HIPS HIPS SEC FT FT/SEC C 0 0.0 0.0 3455.0 0.001740 0.0050 2.04 4.91 SPECIAL BARRIER SPRING O I
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, O LOAD DROP AHALYSIS FOR RCCRr--CASE se a 33.4 FT a 2.7s HIPS O maammmmmmmmmmmmmmmmmmmmmmmmmmmmmmmusummmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmme O DATA CH HISSILE BARRIER, Ate LOAD C0tBINATION EWATION O
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nu mmm mmmmm mmmm mm m m m m mmm mm m mm m m m m m m m m m m mm mmmmm m m m u mmm m m m mm mmmm mm m m m m mm m a n d a m m m m m mm m mm mm mmmm m m m m m mmm mmmmm aammm m m me O .
O BARRIER FORCE DISPLACEHEfR RELATIONSHIP i
HIPS FEET 197.5 0.0424 197.E 0.4240
,. 197.5 1.0000 l O i 0.0 HIPS EQUIVALEHT STATIC FORCE um LOAD 1
'. 0.0 HIPS EQUIVALENT C0ttSTANT DYtIAHIC FORCE um LOAD 2 1 ' O.0 HIP-SEC HISSILE ItPtLSE RESISTED BY FORCE AT BARRIER SUPPORT PLUS BARRIER INERTIA DistPE H0HENTIA4 TRANSFER mm LOAD 3 4' O 3.940 HIP-SEC HISSILE ItPtLSE RESISTED OtLY BY BARRIER INERTIA DURING H0HENTIA4 TRANSFER mm LOAD 4 26.9 FPS BARRIER INITIAL VELOCITY DUE TO LOAD 4 I
f O BARRIER HISSItE HISSItE BARRIER BARRIER BARRIER C EQUIVALENT HEIGHT HEIGHT PLASTIC EFFEC. YIELD PERICO HEIGHT LOAD 3 LOAD 4 FORCE DEFLECTION O HIPS HIPS HIPS HIPS FT SEC C i
1.960 0.000 2.750 197.5 0.0424 0.0227 O C m aam m mmm mmm m mm m mmmmm mm mmm mmm m m m m m m m mm m mmmmmm m m mm mmmmmmm m m m m m m m m m mm mm m m m m m m m m m m m m mm mm m mm m mmmm m mm m un amma mm u na m a n n e RESULTS OF TIHE HISTORY AHALYSIS FOR HISSILE IIPACT HITH OTHER LOADS O C, unamu mm mmmmmmmm mmmmm mmmmmmmm mm m mm mmmmmmm mm m m m m m m m m mm m m m m m m m m mm m mes m u m m m m m m m m m m m m m m m m m m mmmmm m 4 mmm mmmmm mm m m m mm m m m e j
C 1 2 3 4 5 4 7 8 9 C
! TIHE DURATION HISSILE FORCE AT TIHE OF HAX HAXIHUM HAXIHUH HAVIHUH FINAL BARRIER l HISTORY OF FORCE BARRIER BARRIER BARRIER BARRIER BARRIER RESISTING HECHAHISH HUtBER LOAD 3 LOAD 3 SUPPORT DEFLECTION DEFLECTION DUCTILITY VELOCITY O FT/SEC e
_ SEC HIPS HIPS SEC FT Q 0 0.0 0.0 197.5 0.020738 0.2901 4.84 24.94 SPECIAL BARRIER SPRING Q By 2.b.d ATTACHMENT 7 T . 6 9/83 CALC. NO. C-o/ f ilw & q,gn JO 142 3 .17 PAGE 6 0F 7-A b h k l) 7)) a _
as O n ST-331.S8tetI. VERS-00 LEVEL-00 70.I72 13.53.27 C LOAD DROP ANALYSIS FOR RCCR2--CASE 814 58.4 FT & 5.00 HIPS O
aumenna nesee m ama na e m a n um a n um a m m a me m a m em e n n ma a m a na m m m m m m ma a m m m e n s u u m a m a n u m m e s sa n um m mmmm m mmmmm a maam me.ameser- ,emo O DATA CH HISSILE. BARRIER. Ate LOAD C0tBINATION EQUATION O esammeasemenomammaammamenmanosumannunmanammaammeurassumanumannammmmmmmmmmmmmamaannumusmans.mummamanummannumesumum i
O O BARRIER FORCE DISPLACEHENT RELATIONSHIP HIPS FEET g 3455.0 0.0024 3455.0 0.0246 3455.0 0.2000 0.0 HIPS ErWIVALENT STATIC FORCE en LOAD 1 0.0 HIPS EQUIVALENT CONSTANT DYHAHIC FORCE en LOAD 2 0.0 HIP-SEC HISSILE ItG8tA.SE RESISTED BY FORCE AT BARRIER SUPPORT PLUS BARRIER IHERTIA DLNtING H0HENTW TRANSFER me LOAD 3
- O 9.480 HIP-SEC HISSILE IPPULSE RESISTED Ott.Y BY BARRIER INERTIA OURING H0HENTW TRANSFER em LOAD 4 8.4 FPS BARRIER IHITIAL VELOCITY DUE TO LOAD 4 l
O BARRIER HISSILE HISSILE BARRIER BARRIER BARRIER -
EQUlVALEHT HEIGHT HEIGHT PLASTIC EFFEC. YIELD PERIOD HEIGHT LOAD 3 LOAD 4 FORCE DEFLECTION O HIPS HIPS HIPS HIPS rT SEC - C I
! 31.420 0.000 5.000 3455.0 0.0024 0.0052 O .....m.....
C
.........m...m.m...mm.......m..m.a.....m...mmmmmmm.m.........<..m.ma.mmm.....m..a.m..m............m REStA.TS OF TIHE HISTORY ANALYSIS FOR HISSILE IIPACT HITH OTHER LOADS us aa m m m m mm.nen nu mmmm m m ma meuma nu ma n u m ammus u mma m mm m m m m m mmmm m mm m m m a n u n um u u m m u n a n n u ma m m m mm mmmm mm m m mm ui nn ma n u ma na m ma m l
O I TIHE 2
DURATION HISSILE 3 4 5 FORCE AT TIME OF HAX 6
HAXItRJH 7
HAXIFAJH 8
HAXIttAt 9
FINAL BARRIER C
HISTORY OF , FORCE BARRIEW BARRIER BARRIER BARRIER BARRIER RESISTING HECHANISH tAABER LOAD 3 LOAD 3 SUPPORT DEFLECTIOH DEFLECTION DUCTILITY VELOCITY U L
_, SEC HIPS HIPS SEC FT FT/SEC U O 0.0 0.0 3455.0 0.002900 0.0128 5.25 8.38 SPECIAL BARRIER SPRING (,[)
L Pep t 7n. c.A ATTACHMENT 7 C
- c. / g , CALC. NO. c-o/ O 9 JO /03 5./7 @' ,
4@5 PAGE 6 0F 7 D lW b y 1//o
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f .C ST-331.S8tetI.VE'.S-00 LEVEL-00 70.172 13.53.27 O LOAD DROP ANALYSIS FOR RCCR2--CASE 82 3 33.4 FT & 5.00 HIPS O
........ .....................u........m.........u........................mm....m............m..m... .........
O DATA CH HISSILE. BARRIER. Arc LOAD C0tBINATIOH EQUATION *
....... ...m..m....m....m..m...m....m........m....m....m..m...m....m.aum........ .m.au..m.un...i.e...... ......
O O BARRIER FORCE DISPLACEHENT RELATIONSHIP HIPS FEET 197.5 0.0424
, 197.5 0.4240 197.5 1.0000 0.0 HIPS EQUIVALENT STATIC FORCE .. LOAD 1 0.0 HIPS EQUIVALENT CONSTANT DYHAHIC FORCE .. LOAD 2 j 0.0 HIP-SEC HISSILE DG8tR.SE RfSISTED BY FORCE AT BARRIER SUPPORT PLUS BARRIER IMERTIA DL5 TING H0HENTtM TRANSFER .. LOAD 3 _
O 7.170 HIP-SEC HISSILE ItPULSE RESISTED Otr Y BY BARRIER INERTIA DURING H0HENTIM TRANSFER .a LOAD 4 '
! 33.2 FPS BARRIER INITIAL VELOCITY DUE TO LOAD 4 O BARRIER HISSILE HISSILE BARRIER BARRIER BARRIER I.,
EQUIVALENT HEIGHT HEIGHT PLASTIC EFFEC. YIELD PERIOD HEIGHT LOAD 3 LOAD 4 FORCE DEFLECTION O HIPS HIPS HIPS HIPS FT SEC C 1.940 0.000 5.000 197.5 0.0424 0.0227 i
O ..........m...........mm......................m.....m.....u.........mm........m...m............m..............m.
C RESLR.TS OF TIHE HISTORY ANALYSIS FOR HISSILE ItPACT HIIH OTHER LOADS ..
O U
.mmm.......ummm...m.m.....un...m.au.au...un.un.....m...m.un..m..unen.. mum.au..m.au..n...m..m.....m..m....un.....
O 1 2 3 = 5 4 i e , O TIHE DURATION HISSILE FORCE AT TIHE OF HAX HAXIHUH HAXIHUM HAXItRM FINAL PARRIER HISTORY OF FORCE BARRIER BARRIER BARRIER BARRIER BARRIER RESISTItG HECHANISH
< NUISER LOAD 3 LOAD 3 SUPPORT DEFLECTION DEFLECTION DUCTILI1Y VELOCITY U HIPS FT/SEC
,, SEC HIPS SEC FT J
V 0 0.0 0.0 197.5 0.036934 0.6234 14.49 33.17 SPECIAL BARRIER SPRING V U '
NcD EA ATTACHMENT 7
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i fCHO 1 2 3 4 5 4 7 8 O 12345478901234567890123954789012345470901234547690123454789012345478901234547890 LOAD DROP ANALYSIS FOR FHCR7-AHALYSIS BY- R.D. SALTER...CHECHED BY ....... 1 F
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O DATA CH HISSILE. BARRIER. Ate LOAD COPSINATION EQUATION
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4.4 FPS BARRIER INITIAL VELOCITv DUE TO LOAD 4 O sARRIER HISSItE HISSILE nARRIER nARRIER nARRIER C i EQUIVALENT HEIGHT HEIGHT PLASTIC EFFEC. YIELD PERIOD d
HEIGHT LOAD 3 LOAD 4 FORCE DEFLECT 10H i '
O HIPS HIPS HIPS HIPS rT SEC C 47.700 0.000 10.000 3314.0 0.0244 0.0244
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RESLLTS OF TIME HISTORv ANALvSIS FOR HISSILE Itt'ACT HITH OTHEW LOADS
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l C 1 E 3 4 5 6 7 8 9 FINAL BARRIER C
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TIHE DURATION HISSILE FORCE AT TIME OF HAX HAXIHUM MAXIHUM HAXIHUN l
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. j O O 0.0 0.0 2531.7 0.004560 0.0105 0.75 4.39 SPECIAL BARRIER SPRING O
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fC LOAD DROP HYDRALA.IC JACH DROP DN FUEL POOL MALLS eseeeeeeeeeeeeeeeen..eeeeeeee.e..eeeeeee.een.eeeeeeeee...... ..eeen.e............e.eeeeeeeeen..eeeeeeeee.eeeeeee O DATA DN HISSILE, sARRIER. Are LOAD C0teINATsaN EQUATIDH O
.eeeeeeeeeeeeeee.ees.eee.*ee.....e.eeeeeee.e.ne..mee.e.neesemes........see.... eeeeee..eeeeeees.eeeeeeeeeeeeeeen i :l O O I BARRIER FORCE DISPLACEHENT RELATIONSHIP l MIPS FEET lO C i 0032.0 0.0307 0032.0 0.3073 8032.0 1.0000 l O 8.0 MIPS Equ1 VALENT STATIC FORCE .. LOAD 1 C
i 1 0.0 HIPS EQUIVALENT CONSTANT DYHAHIC FORCE .. LOAD I 0.0 MIP-SEC HISSILE IWLA.SE RESISTED BY FORCE AT BARRIER SUPPORT PLUS BARRIER INERTIA DL5 TING MOHENTIM TRANSFFR em LOAD 3 3
! O 10.130 uIP-SEc HISSILE IwutSE RESISTED Ore v av eARRIER INERTIA DuRING MOHEwTLm TRANSrER .. LOAD = C
, t 10.2 FPS BARRIER INITIAL VELOCITY DUE TD LOAD 4
- 1 1
iO nARRIER EQUIVALENF HISSItE HEIGHT MISSItE HEIGHT sARRIER nARRIER sARRIER PLASTIC EFFEC. YIELD PERIOD C
- t. HEIGHT LOAD 3 LOAD 4 FORCE DEFLECTION O uIPS HIPS uIPS HIPS rT SEC 'l I 24.000 0.000 7.000 8032.0 0.0307 0.0112
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l RESLA.TS OF TIME HISTORY ANALYSIS FOR HISSILE Ite'ACT HITH OTHER LOADS ,.
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- s 4 7 3 0 0 TIHE DURATION MISSILE FORCE AT TIHE OF HAK HAXIHUM HAXIHUM MAXIHUH FINAL BARRIER HISTORY OF FORCE BARRIER BARRIER BARRIER BARRIER BARRIER RESISTING HECHANISH HLReER LOAD 3 LOAD 3 SUPPORT DEFLECTIDH DEFLECTION DUCTILITY VELOCITY SEC NIPS HIPS SEC FT FT/SEC i
O O 0.0 0.0 8032.0 0.003300 0.0372 1.21 18.17 SPECIAL BARRIER SPRING C ATTACHMENT 0) ='
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O DATA OH HISSItE. sARRIER Ate LOAD COteINATION EQUATION r O C BARRIER FORCE DISPLACEMENT RELATIONSHIP HIPS FEET 400.0 0.0024 408.0 0.0240 408.0 1.0000 O.0 HIPS EQUIVALENT STATIC FORCE .. LOAD 1 0.0 HIPS EQUIVALENT CONSTANT DYHAHIC FORCE .. LOAD E i 8.0 HIP-SEC HISSILE DFLA.SE RESISTED BY FORCE AT BARRIER SUPPORT PLUS BARRIER INERTIA DLARIHS H0HENTW TRANSFER .. LOAD 3 '.~
O 10.130 HIP-SEC HISSILE DruLSE RESISTE0 Ota Y av eARRIER INERTIA DuRIHs H0HENTiAt TRe2<SrER .. LOAD =
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O sARRIER EQUIVALENT HISSItE HEIGHT HISSItE HEIGHT sARRIER eARRIER PLASTIC EFFEC. YIELD nARRIER PERIOD HEIGHT LOAD 3 LOAD 4 FORCE DEFLECTION
- O HIPS HIPS HIPS HIPS FT SEC (
13.250 0.000 7.000 408.0 0.0024 0.0080 0 C RESLA.TS OF TIHE HISTORY ANALYSIS FOR HISSILE DFACT HITH OTHER LOADS j ... . .. ........................................................................................ .....
O 1 2 3 4 5 6 7 8 9 O TIHE DURATION HISSILE FORCE AT TIHE OF HAX HAXDRAt HAXDeA3 HAXIHUH FINAL 6ARRIER
- HISTORY OF FORCE BARRIER BARRIER BARRIER BARRIER BARRIER RESISTING HECHANISH l HUPBE.1 LOAD 3 LOAD 3 SUPPORT DEFLECTI0tl DEFLECTION DUCTILITY VELOCITY U O I
SEC HIPS HIPS SEC FT FT/SEC O O 0.0 0.0 40s.0 0.03I404 0.4707 100.as 30.42 SPECIAL BARRIER SPRING C i
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- 6. Head and Internals Handling Fixture Assembly ;
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