ML20049H856
| ML20049H856 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 02/26/1982 |
| From: | Jackie Cook CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR 15979, NUDOCS 8203040368 | |
| Download: ML20049H856 (150) | |
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i Pr sil - Projects, Engineering ~ and Construction General offices: 1945 West Parnal! Road, Jackson, MI 49201 e (517) 788 0453 February 26, 1982 80 e s Office of Nuclear Reactor Regulation ( Skc k b Harold R Denton, Director Division of Licensing -N O US Nuclear Regulatory Commission (U%gr 8pg [ Washington, DC 20555 MIDLAND PROJECT D MIDLAND DOCKET NOS 50-329, 50-330 m PART II RESPONSE TO NUREG-0612 CONTROL OF HEAVY LOADS AT NUCLEAR POWER PLANTS FILE: 0963 SERIAL: 15979 Reference (A) J W Cook letter to H R Denton, Serial 14928, Dated December 21, 1981 Enclosure (1) Midland Units 1 & 2 - Part II Response - Control of Heavy Loads at Nuclear Power Plant (NUREG-0612) (2) Updated Pages 9, 10, Table 3, Sheet 2, and Table 4, Sheets 1-3, of Part I Response - Contcol of Heavy Loads at Nuclear Power Plants (NUREG-0612) Reference A submitted Part I Response for control of heavy loads and stated that Part II Response would be submitted to the Staff by February 26, 1982. The results of CP Co's Part II review are enclosed and address the specific requirements for the control of heavy loads in spent fuel pool, reactor vessel and other safe shutdown areas as outlined in D G Eisenhut's letters of December 22, 1980 and February 3, 1981. The "C" series drawings in Part II response supersede those from Part I. Analyses of heavy load drops onto the auxiliary building roof and of core flood line integrity af ter a RV head drop onto the RV are still underway. We expect these analyses to be completed in the near future and will forward the results as soon as they become available. We believe our Part I and Part II responses satisfactorily address the con-cerns of NUREG-0612. fd si io q, F L oc0282-0693a131 82"3040368 820226 PDR ADOCK 05000329 A PDR
i 2 J Also enclosed are revised pages 9, 10 and revised sheets of Tables 3 and 4 from our Part I Response. These revised pages complete the follow-up commitments made in the Part I response. JWC/ PEP /fms CC RJCook, Midland Resident Inspector DBMiller, Midland (3) RHernan, US NRC RWHuston, Washington i LSRubenstein, US JRC l-T J i I ~ l t f t i i f. I i t f f. P P oc0282-0693al31 ---v-r< ,,-r ,w,ysm .-y e- - r v e y
..e..- 9 Item 3d (2.1.3d) Verification that lifting devices identified in 2.1.3c, above, comply with the requirements of ANSI N14.6-1978, or ANSI B30.9-1971 as appropriate. For lifting devices where these standards, as supplemented by NUREG-0612, Section 5.1.1(4) or 5.1.1(5), are not met, describe any proposed alternatives and demonstrate their equivalency in terms of load-handling reliability. Response - Non-special lift devices (sling, shackles, etc) purchased are required to meet the guidelines of ANSI B30.9-1971 as specified in the purchase specification. These devices are marked with load limits and will be inspected before use for heavy load lifts. The spent fuel cask lifting device to date has not been purchased. It is CP Co's intention to have the spent fuel cask handling system meet the single-failure proof guidelines of NUREG-0612, Section 5.1.6, therefore, the handling device will also be specially designed and built to the single-failure proof criteria of NUREG-0612, Section 5.1.6. Following is a brief description of the special lifting devices. 1. Reactor Vessel Head and Internals Handling Fixture - Tripod configured carbon steel weldment designed for handling the reactor vessel closure head assembly and the reactor vessel internals. 2. Internals Handling Extension - A carbon steel assembly designed to connect the head and internals handling fixture to the main hook of the polar crane. 3. Internals Handling Adapter - An assembly designed for handling the plenum assembly, the core support assembly, or the core support assembly containing the plenum assembly. This adapter is used in 4 conjunction with the head and internals handling fixture, the internals handling extension, and the internals indexing fixture for internals handling. 4. Stud Tensioner Sling - a four-legged sling used for moving each stud tensioner from its storage location into the transfer canal area and for transferring the load of the tensioner from the reactor building crane to the stud tensioner hoist. 5. Fuel Transfer Carriage Lifting Rig Assembly - used to lift the fuel transfer carriage. 6. New and Failed Fuel Handling Tool and Sling - used to lift the New Fuel Assemblies and the Failed Fuel Container. 1 7. Quad Sling - used to move the filter. handling machine. 8. Spreader Beam Type IA - used for handling decontamination area access hatch and spent resin area access hatches.
.. _ _ _. _ _ _. - - -. _ _ _ _ _ ~. _. _.. _. ~. _ _. _ _ _ _. _, 10 9. Spreader Beam Type IB - used for handling filter plugs, demineralizer plug, degasifier plugs and pipe floor chase shield plugs. 10. Spreader Beam Type II - used for handling spent resin area access hatches. All of the special devices listed above were designed prior to the exis-tence of ANSI N14.6-198 and NUREG-0612. They were designed in accordance with accepted industry standards and sound engineering practices. Devices 1 through 4 above were designed with a safety factor of three based on yield strength and considering static load. In addition, an initial acceptance test of 150% of static load has been performed. Device 5 was designed with a safety factor of one. Device 6 was designed with a safety factor of 3. Device 7 was designed in accordance with ANSI B30.9-1971 1 with a safety factor of 1.5 and was tested to twice the actual load. Devices 8, 9 and 10 design included 25% impact load and the resulting safety factors based on the limiting component of the assembly are 1, 1.5 and 1.2 respectively. All the special lifting devices will be maintained and inspected in accordance with the requirements of ANSI N14.6-1978, Sections 5.1, 5.3, 5.4 and 5.5. The non-special lifting devices will be inspected and maintained in accordance with the requirements of ANSI B30.9-1971. For these reasons, we believe that these devices are acceptable for use at Midland.
2 Trble 3 (Continued) Load Weight Load Path HandlinI3) Load Name (tons) llandling System Lifting Device Figure Number Procedure Boom Crane 3.6 Reactor Building Crane C-16, C-15 Auxiliary Building Crane C-15 L Letdown Cooler 3 Rigging Beam C-2 Reactor Building Crane C-16 Auxiliary Building Crane C-7, C-13 Seal' Plate 6 Reactor Building Crane C-5 Missile Shields 52.5 Reactor Building Crane C-16 (4 per unit) Equipment Carriage Reactor Building Crane C-7, C-5 i Auxiliary Building Crane C-7, C-13 } l Plentun Assembly and Core 180 Reactor Building Crane See Upper C-6 Support Assembly Internals Reactor Vessel Top Head 3.8 Reactor Building Crane C-5, C-17 l Insulation Rack Assembly 1 Internals Indexing Fixture 6.5 Reactor Bualding Crane C-9(5) li g l l t i \\ (1)lleavy loads transported through both the reactor and auxiliary buildings indicated by listing handling systems in both buildings (2) Transported to and from reactor building in equipment carriage (3) Procedures not written to date; every heavy load lift will be covered by a procedure (4)Not procured at this time; therefore, weight is unknown (5) Path to be added in Part. II response.
i i TABLE 4 L HEAVY LOADS ( AUXILIARY BUILDING ( } l UNITS 1 AND 2 j i l Load Weight Load Path Handlin . +i Load Name (tons) Handling System Lifting Device Figure Number Procedure {2) i Spent Fuel Shipping Cask 15-110 Auxiliary Building Crane C-9, C-13 j Neutron Source Shipping Cask 12 Auxiliary Building Crane C-9, C-13 Irradiated Specimen Shipping 3.5-12 Auxiliary Building Crane C-9, C-13 Cask New Fuel Shipping Containers 3-4 Auxiliary Building Crane C-11, C-13 Failed Fuel Container ( 1 Auxiliary Building Crane New & Failed 1 Fuel Handling Tool & Sling Fuel Transfer Carriage 2.5 Auxiliary Building Crane Fuel Transfer C-10, C-13 Carriage Lifting Rig Assembly Crane Load Block (Main) 5.8 Auxiliary Building Crane A-5, A-6, B-7, B-8 Boom Crane 3.6 Auxiliary Building Crane C-15, C-16 l Makeup Pump, 3 per unit 3 Makeup Pump Hoist C-2 Auxiliary Building Crane C-2 to.C-6, C-13 Auxiliary Feedwater Pump 2.8 Auxiliary Feedwater Pump C-1 i Hoist j Auxiliary Building Crar.e C-1 to C-6, C-13
2' I T:ble 4 (Continued) f Handlinf2) ( Load Weight Load Path Load Name (tons) Handling System Lifting Device Figure Number Procedure I t Main Steam Isolation Valve 7.5 Auxiliary Building C-18 Operator Electric Monorail Auxiliary Building Crane C-8, C-13 Process Steam Transfer Valve 7 Auxiliary Building C-18 Operators Electric Monorail i Auxiliary Building Crane C-8, C-13 l 1 Filter Handling Machine 5.5 Auxiliary Building Crane Quad-Sling C-12, C-13 l Filter Transfer Cask 2.6 Auxiliary Building Crane C-12, C-13 Decontamination Room Jih C-4 Crane Equipment Access Hatch 1.25 Auxiliary Building Crane C-1, C-14 El 584'-0" Equipment Access Hatch 1.25 Auxiliary Building Crane C-2, C-14 'El 599'-0" Equipment Access Hatch 1.25 Auxiliary Building Crane C-3, C-14 El 614'-0" Equipment Access Hatch 1.25 Auxiliary Building Crane C-4, C-14 El 634'-6" i Equipment Access Hatch 1.25 Auxiliary Building Crane C-5, C-14 El 645'-0" Equipment Access Hatch 1.25 Auxiliary Building Crane C-14 El 659'-0" Decontamination Area 6.9 Auxiliary Building Crane Spreader Beam - C-15 1 l Access Hatch Plug No 1 Type IA (2 pieces)
I Table 4 (Continued) Load Weight Load Path llandlinI2) Load Name (tons) llandling System Lifting Device Figure Number Procedure Filter Plug No 2 6.8 Auxiliary Building Crane Spreader Beam - C-14 l1 (24 places) Type IB Spent Resin Access llatch El 659'-0" Plug No 3 (2 pieces) 24 Auxiliary Building Crane Spreader Beam - C-15 I Type IA and II Plug No 4 (2 pieces) 20 Auxiliary Building Crane Spreader Beam - C-15 1 Type IA and II Demineralizer Plug No 5 11.5 Auxiliary Building Crane Spreader Beam - C-14 (13 places) Type IB I Degasifier Plug No 6 11.7 Auxiliary Building Crane Spreader Beam - C-14 (2 places) Type IB Pipe Chase Access Plugs El 659'-0" I Plug No 7 (2 places) 2.2 Rigging Beam lloist C-15 I Plug No 8 1.9 Rigging Beam floist C-15 lI Pipe Floor Chase 3.9 Portable-Spreader Beam - C-4 El 634'-6" Type IB i Shield Plug No 9 l (2 places) l Valve Pit - Steel Deck 1.1 Portable C-1 El 584'-0" (2 places) B
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l l CONSUMERS P0kIR COMPANY MIDLAND UNITS CONTROL OF HEAVY LOADS AT NUCLEAR P0kTR PLANTS (NUREG-0612) PART II RESPO"SE I
i CONTROL OF HEAVY LOADS AT NUCLEAR POWF". PLANTS PART II RESPONSE TABLE OF CONTENTS Palle Specific Requirements for Overhead Handling Systems Operating in Vicinity of Fuel Stotage Pools (2.2) 1 Heavy Load Systems (2.2.1) 1 Exclusion (2.2.2) 2 Single-Failure-Proof Systems (2.2.3) 3 Non-Single-Failure-Proof Systems (2.2.4) 6 Alternatives (2.2.4a) 6 Crane Motion Limitation (2.2.4b) 7 Crane Operational Limitation (2.2.4c) 8 U Physical Separation Between Spent Fuel and Heavy Loads (2.2.4d) 9 Analyses for Compliance with Criteria I through III (2.2.4e) 10 Specific Requirements of Overhead Handling Systems Operating in the Containment (2.3) 11 Heavy Load Systems (2.3.1) 11 ) Exclusion (2.3.2) 12 Single-Failure-Proof Systems (2.3.3) 13 Non-Single-Failure-Proof Systems (2.3.4) 14 Electrical Interlocks and Mechanical Stops (2.3.4a) 14 Other Site-Specific Considerations (2.3.4b) 15 Analyses for Compliance with Criteria I through III (2.3.4c) 16 l 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 (2.4) 20 ' Single-Failure-Proof Systems, All Loads'(2.4.1) 20
11 (3 v' TABLE OF CONTENTS (Continued) Page Non-Single-Failure Proof Systems (2.4.2) 21 Heavy Loads and Potential Impact Areas (2.4.2a) 21 2 Load / Impact Area Elimination (2.4.2b) 22 Elimination Because of Redundancy and Separation (2.4.2b.1) 22 Elimination Because of Mechanical Stops or Electrical Interlocks (2.4.2b.2) 22 Elimination Because of Other' Site-Specific Considerations (2.4.2b.3) 22 Single-Failure-Proof System, Specific Loads (2.4.2c) 23 Analysis for Compliance With Criterion IV (2.4.2d) 24 Load Retention During SSE (2.4.2d.1) 24 Exception to Analytical Guidelines of NUREG-0612, Appendix A (2.4.2d.2) 24 Information (2.4.2d.3) 24 List of Tables Table 1 - Load / Safe Shutdown Area Matrix for Reactor Building Table 2 - Load / Safe Shutdown Area Matrix for Auxiliary Building List of Figures C-1 Load Paths at El 584'-0" C-2 Load Paths at El 599'-0" C-3 Load Paths at El 614'-0" C-4 Load Paths at El 634'-6" C-5 Load Paths at El 645'-0" C-6 Load Paths at El 659'-0" /5 C-7 Load Paths at El 659'-0" (,_ C-8 Load Paths at El 659'-0" l
iii - TABLE OF CONTENTS (Continued) C-9 Load Paths at El 659'-0" C-10 Load Paths at El 659'-0" C-11 Load Paths at El 659'-0" C-12 Load Paths at El 659'-0" C-13 Load Paths at El 659'-0" and 634'-6" C-14 Load Paths at El 659'-0" C-15 Load Paths at El 659'-0" C-16 Load Paths at El 685'-0" ~: C-17 Load Paths at El 685'-0" C-18 Load Paths at El 704'-0" Appendices Appendix A Supplementary Information for Off-Site Radiological Release Analysis - Criterion I, Reactor Building Appendix B Reactor Vessel Head Drop Analysis - Criterion III, Reactor Building Appendix C Supplementary Information for Load Drop Analysis - Criterion IV, Auxiliary Building (El 659'-0") _i / f, i 1 Y f '4 r g.
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CONTROL OF HEAVY LOADS AT NUCLEAR POWER PLANTS PART II - RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON " CONTROL OF HEAVY LOADS" The following report provides Consumers Power Company's Part II response to the NRC's request for additional information on the control of heavy loads. Part I response was submitted to the NRC by J W Cook letter Serial 14928 dated 12/21/81. The items identified in parenthesis below refer to those outlined in Enclosure 3, Section 2.1 of NRC letters dated December 22, 1980 and February 3, 1981, D G Eisenhut to All Licensees of Operating Plants and Applicants for Operating Licenses and Holders of Construction Permits. 2.2 Specific Requirements for Overhead Handling Systems Operating in the Vicinity of Fuel Storage Pools Item I (2.2.1) Identify by name, type, capacity, and equipment designator, any cranes physically capable (ie, ignoring interlocks, moveable mechanical stops, or operating procedures) of carrying loads which could, if dropped, land or fall into the spent fuel pool. {'~'/ Response - Handling systems capable of carrying loads over the spent fuel S g_, pool as shown in the "A" series drawings of our Part I response include: 1. Auxiliary Building Crane (OH-52) - An Ederer rectilinear crane with a main hoist of 125 ton capacity and an auxiliary hoist of 15 ton capacity. 2. Auxiliary Building Fuel Handling Bridge (OH-53) - Capacity 1.5 tons (see Response to 2.2.4a below). 3. Auxiliary Building Fuel Handling Bridge Monorail - Capacity 1.5 tor,s and connected to the fuel handling bridge (see response to 2.2.4a below). 4. Auxiliary Building Electric Monorail - Capacity 7.5 tons - Used to handle main steam isolation valve and process steam valve operators. A f O :,- ,o 1
g s f -s v' 2 i o Item 2 (2.2.2) e 'b ). Justify the exclusion of any cranes in this area from the above category j' (2.2.1) 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 1 demonstrating that for any failure mode, no heavy load can fall into the fuel storage pool. ) 4 Response - The remaining handling sy' stems in the auxiliary building have i i been excluded from 2.2.1, above, because they are not in close proximity j to the spent fuel pool. The handling systems excluded are: h 1. Auxiliary Feedwater Pump Hoists 1 s a '.\\ x l,, 2. Makeup Pump Hoists 3. Decontamination Room Jib Crane i 1 4. Boom Crane 5. Filter Plug Hoist N \\ g 6. Rigging Beam Shield, Plugs. 1 J j I 1 l I f i i 4 g s s Y, h ~ 4. e .g 4 \\ A w .,w
3 O V Item 3 (2.2.3) Identify any cranes listed in 2.2.1, above, which you have evaluated as having sufficient design features to make the likelihood of a load drop l extremely small for all loads to be carried and the basis for this evaluation (ie, 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 l system (ie, crane load combination) information specified in Attachment 1. Response - The spent fuel cask load lifted with the main hoist of the auxiliary building crane will have a handling system that meets the single failure proof guidelines of NUREG-0612. Attachment 1 information for this system follows: 1. The auxiliary building crane is an Ederer rectilinear X-SAM crane. The design rated load (DRL) of the main hoist is 125 tons. 2. The main hoist is identical to the X-SAM system described in Ederer Topical Reports EDR-1(P)-A and EDR-1(NP) n with the exception that the Midland crane has a single load path lower hook (ie, trunnions and attaching points). Because of lifting envelop limitations, the Midland design was not able to accommodate dual load path attaching (,, points. In view of this limitation, a design factor of 10:1 has been applied to the trunnion and attaching point components that have a single load path. The Ederer topicals were reviewed and approved by the NRC as documented in NRC letter R L Baer to C W Clark dated January 2,1980. 3. The methods and assumptions employed in the seismic analysis to demonstrate that the Midland X-SAM system can retain a load during a SSE include: A. To demonstrate that the main hoist of the auxiliary building crane can retain its load during a safe shutdown earthquake an evaluation was made utilizing the FORTRAN program BCDAWS/1. B. For the seismic analysis the following assumptions and mathematical model were employed:
- 1) Vertical Direction J
a) Only the first flexural mode of girder is considered b) Lumped masses c) Vertical motion is uncoupled from either direction d) 2% damping x_ -
4 V Only the first flexural mode is considered because subsequent modes are assumed to have a negligible effect. Frequencies and participation factors are calculated by means of classical eqt.ations and accelerations are determined from Midland Seismic Specification #G-7. A damping value of 2% was used for both OBE and SSE. Although the ends of the crane bridge were assumed rigid in the initial model, the flexibility of the crane rail girder was considered in a second model. The assumption of rigid supports was shown to be conservative.
- 2) Longitudinal Direction f
a) Longitudinal motion is uncoupled from other directions b) Hook load is omitted i c) 2% damping Motion is analyzed as a single degree of freedom system. Hook load is omitted due to the low frequency of pendulum motion. Calculations show that the first two modes are virtually uncoupled. For this case, the crane bridge is rigid with respect to the supporting rails. Stresses due to [] longitudinal seismic motion are extremely small (less than (s 300 psi) and, although these stresses may increase slightly due to support girder flexibility, they will still be negligible.
- 3) Lateral Direction a) Lateral motion is uncoupled from other directions Lateral seismic stress is calculated assuming slip along rails occurs under the most conservative conditions, that being maximum upwards acceleration. Stress is determined frcm the maximum vertical reactions and the coefficient of friction. Sill beam stiffness was considered in the analysis and had the effect of reducing stresses induced by seismic forces lateral to the girder.
l C. The hook and trolley position was selected by trial and error. i Several hook and trolley position combinations were evaluated for OBE and SSE load cases. i i Cable Length 69" (High hook) = 909" (Low hook) = Trolley Position 34.12' (Mid span) = = 17.06' (Quarter span) = 4.0' (Sixteenth span - extreme position) O ) It was determined that maximum crane girder stress occurs with high hook and trolley at mid span.
)' 5 D. Summary - the maximum girder stress is.69 Fy for the OBE condition and.81 Fy for the SSE condition. Allowable girder stress is.90 Fy for SSE. No allowable stress is given for OBE. 4. The lifting device for the spent fuel cask has not been purchased to date. CP Co anticipates there will be a need for this device approximately five years after commercial operation. It is CP Co's intention that this device be designed and built to the criteria of NUREG-0612, Section 5.1.6. 5. The spent fuel cask has not been purchased to date. When it is procured, CP Co intends that the interfacing lift points for the cask used at Midland, will meet the guidelines of NUREG-0612, Section 5.1.6. + 1 e l l t i 1 1 f i - 1 J I 1 i 7 .=-r..- .cr-,-w r y v-y- ..v---v--vv .--1-w-g-- ,~ m p--. =, -,,, - - -,, - --.7 -c,- -+
6 bv Item 4 (2.2.4) 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 Criteria 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 operation in the spent fuel area and your determination of compliance. This response should include the following information for each crane. Item 4a (2.2.4a) Which alternatives (eg, 2, 3 or 4) from those identified in NUREG-0612, Section 5.1.2, have been selected. Response - The auxiliary building fuel handling bridge does not handle heavy loads, therefore, the criteria of NUREG-0612, Section 5.1.1, is not applicable. Besides handling fuel this system lifts control components and fuel inspection system components. The auxiliary building fuel handling bridge monorail is used to open the spent fuel pool gates which are not considered a heavy load. Y (w]
7 Item 4b (2.2.4b) 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 specifica-tion (operational and surveillance) provided to ensure the operability of such electrical interlocks or mechanical stops. Response - The auxiliary building crane, although capable of handling heavy loads, is equipped with mechanical stops which prohibit any load handled with this crane from passing over the spent fuel pool. No heavy load lifts will be made over a fuel pool containing spent or new fuel. The closest a heavy load can come to the spent fuel pool is approximately 6 feet. This distance is reached for the spent fuel cask, neutron source shipping cask and the irradiated specimen shipping cask lifts (load path Figure C-9); and for the fuel transfer carriage lift (load path Figure C-10). The physical size of these loads, with the exception of the spent fuel cask, are such that, in the unlikely event of the load dropping and tipping, the load would not reach the spent fuel pool. The spent fuel cask lift will be handled with a single-failure proof system. / (s,-} The auxiliary building electric monorail can carry loads over the spent fuel pool (see Part I Response Figure A-9). Figure C-18 shows the desired load path for heavy loads handled with the monorail. The procedure (s) used to handle heavy loads with this monorail will require the installation of mechanical stops on the branches that travel over the spent fuel pool (Figure A-9 coordinates 5.6 and 7.4, Part I Response). These stops will ensure the hoist will not travel over the spent fuel pool when handling heavy loads and they will be removed after the lift is completed. For the reasons given above, CP Co believes that the intent of NUREG-0612, Section 5.1.2, has been met and no further analyses are necessary. I \\ \\ x_ /
l \\ 8 O Item 4c (2.2.4c) Where reliance is placed on crane operational limitations with respect to f 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 physical controls provided to ensure that these assumptions remain valid. Response - The Midland Plant analysis of the spent fuel pool area does not rely on crane operational limitations with respect to the time of the storage of certain quantities of spent fuel at specific post-irradiation decay times, t 1 i i 4 1 O i i --,,.y-. ..7,- ,-._,, _.,,,,,,.,.+. ,,,,,..,-,n.,
t i 9 O Item 4d (2.2.4d) i khere reliance is placed on the physical location of specific fuel modules at certain post-irradiation decay times, provide present and/or proposed technical specifications and discuss administrative or physical controls provided to ensure that these assumptions remain valid. j Response - The Midland Plant analysis of the spent fuel pool area does not rely on the physical location of specific fuel modules at certain post-irradiation decay times. 1 1 p 3 l 1 i l 1 e i l 1 ,,,,.,---,-,,,,.,-,--,------,---,--.--.--,.,,--.r- -,c s,..,,n,- ,--,,,,--n-,
i j j 10 i Item 4e (2.2.4e) ) I Analyses performed to demonstrate compliance with Criteria I through III i should conform to the guidelines of NUREG-0612, Appendix A. Justify any j exception taken to these guidelines, and provide the specific information requested in Attachment 2, 3 or 4, as appropriate, for each analysis performed. Response - Since heavy loads are not lifted over the spent fuel pool, analyses to demonstrate compliance with Criteria I through III are not required. i i i 3 e i i i} f I 4 1 i 't i 4 3 L i + ..++----,----,-----...-...,,-----w, ,,,w.,----w-,v<,,,my---,,-,-,,,--ew.ry.,-- ,w- -,-,+~ev.r,
11 l O 2.3 Specific Requirements of Overhead Handling Systems Operating in the Containment Item 1 (2.3.1) Identify by nama, type, capacity, and equipment designator, any cranes physically capable (ie, taking no credit for any interlocks or operating procedures) of carrying heavy loads over the reactor vessel. 3 Response - Handling systems capable of carrying heavy loads over the reactor vessel include: 1. Reactor Building Polar Crane (1H-51, 2H-51) - Manufactured by Harnischfeger Corporation with a main hoist of 190 ton capacity and an auxiliary hoist of 25 ton capacity. 2. Reactor Building Fuel Handling Bridge (1H-54, 2H-54) - Capacity 1.5 tons (see Response to 2.3.4c below). 3. Boom Crane (OH-20) - Capacity 2 tons ) 1 J O U i r n,.-.-, -n-. .-r--- ,~,- -.. - - -,. - =
12 Item 2 (2.3.2) 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 - The remaining handling systems in the reactor building have been excluded from 2.3.1, above, because as shown in the "A" series drawings of our Part I response they are not in close proximity to the reactor vessel. The handling systems excluded are: 1. Rigging Beam - Reactor Vessel Head Studs 2. Rigging Beam - Snubbers 3. Rigging Beam - Miscellaneous Equipment (Unit 2 Only) 4. Rigging Beam - Letdown Cooler 5. Stud Tensioner Hoists *
- The stud tensioner hoists, although used in close proximity to the
[ \\ reactor vessel, are only in use to preload/ unload reactor vessel studs \\~- and do not handle loads over an open reactor vessel. i \\' m
13 Item 3 (2.3.3) 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 the basis for this evaluation (ie, 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 (ie, crane-load-combination) information specified in. Response - Midland's cranes identified in 2.3.1 do not take credit for single failure proof handling of all loads. 1 L) ,,.,y,
1 14 [ \\ Item 4 (2.3.4) For cranes identified in 2.3.1, above, not categorized according to 2.3.3, demonstrate that the evaluation criteria of NUREG-0612, Section 5.1, are satisfied. Compliance with Criterion IV will be demonstrated in your response to Section 2.4 of this request. With respect to Criteria I through 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 case. Item 4a (2.3.4a) .Where reliance is placed on the installation and use of electrical interlocks or mechanical stops, indicate the circumstances under which these protective devices crn be removed or 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 - The Midland Plant does not rely on the installation and use of electrical interlocks or mechanical stops for cranes carrying heavy loads in the reactor building. l 1 i l s
l-15 s Item 4b (2.3.4b) Where reliauce is placed on other, site-specific considerations (eg, refaeling sequencing), provide present or proposed technical specifications and discuss administrative or physical controls provided to ensure the continued validity of such considerations. Response - The Midland Plant does not rely on other, site-specific considerations (eg, refueling sequencing) for heavy load drop analyses on the reactor vessel. For heavy load drop analyses elsewhere in the reactor building, see Section 2.4 of this report. j l \\
16 !V Item 4c (2.3.4c) Analyses performed to demonstrate compliance with Criteria I through III should conform with the guidelines of NUREG-0612, Appendix A. Justify any exceptions taken to these guidelines, and provide the specific information requested in Attachment 2, 3, or 4, as appropriate, for each analysis performed. Response - The reactor building fuel handling bridge does not handle heavy loads and, therefore, does not warrant further consideration in this analysis. In containment the boom crane is used only to service the control rod drive mechanisms, which are not heavy loads. The boom crane, therefore, does not warrant further consideration in this analysis. Heavy loads that are handled over the reactor vessel include the reactor vessel head, the internals indexing fixture, the upper internals, the in-service-inspection tool and the missile shields. Results of analyses performed for comparison with Criteria I through III follow: Comparison With Criterion I Criterion I - Release of radioactive material that may result from damage to spent fuel based on calculations involving accidental dropping of a postulated heavy load produce doses that are well within 10 CFR Part 100 limits of 300 rem thyroid, 25 rem whole body (analyses should show that N-doses are equal or less than 1/4 of Part 100 limits). Analysis - As a starting point, we have analyzed the consequences of a fuel handling accident. Design basis, conservative, and realistic analyses, as described in FSAR Subsection 15.7.4, were performed to calculate doses received: (1) at the exclusion area boundary during a two hour period immediately following an accident; and (2) at the outer boundary for the duration of the accident. The results are given in the table below: 4 OV ..r.-
17 Design Basis Conservative Realistic Containment Value Value Value Exclusion Area Boundary Dose (0 to 2 hours), rem Thyroid 2.13 .575 7.22E-03 Skin .730 .196 9.76E-03 Whole-Body Gamma .270 7.26E-02 2.65E-03 LPZ Outer Boundary Dose (Duration), rem Thyroid .404 .109 1.03E-03 Skin .138 3.71E-02 1.39E-03 Whole-Body Gamma 5.10E-02 1.37E-02 3.79E-04 Further details of the accident analysis including information requested by Attachment 2 can be found in the Midland FSAR Subsections 15.7.4, 9.4.9, 12.3.4.2, Tables 15.7.5, 15.7.7 and Q&R 312.44. This material is included as Appendix A of this report. As can be seen by the results, the number of fuel assemblies that have to be damaged to exceed 1/4 10 CFR 100 limits at the exclusion boundary would be approximately 85 to 2350 depending on conservative or realistic results. The probability of damage to the fuel from a drop of a heavy load is extremely unlikely. Damage due to a missile generated by a heavy t load drop (RV head guide stud) is more credible but still unlikely consi-dering damage to approximately 1/2 of the 177 fuel assemblies would be required before 1/4 of the 10 CFR 100 limits are exceeded if conservative results are used. CP Co believes that the probability of this event is extremely low and therefore considers that the off-site release criteria is satisfied. Comparison with Criterion II Criterion II - Damage to fuel and fuel storage racks based on calcula-tions involving accidental dropping of a postulated heavy load does not result in a configuratios of the fuel such that K,gg is larger than 0.95. Analysis - NUREG-0612, Appendix A, provides an option for a neutronics analysis. Either demonstrate that the maximum K is no greater than ff 0.95 as shown by a plant specific analysis; or d,monstrate that K for e an uncrushed core is no greater than 0.90. TheNRCestimatesthalgkhe bounding case reactivity increase due to a crushed core is 0.05. O ^
18 bv For the latter option and with the following initial conditions one can calculate a K of 0.838 for the core during refueling, which is well withinthe0.Ibflimit: Refueling Boron Level (FSAR Subsection 16.3.1.1.1) 2270 ppm Boron Worth, 70F, Zero Power (FSAR Table 4.3.12) 1/74% K/K/ ppm Negative Reactivity Due to Boron 30.68% K/K Total Rod Worth (EOL) (FSAR Table 4.3.3) 9.48% K/K K,gg, Cold 70F, (BOL) (FSAR Table 4.3.10) 1.24 In addition to the above, Generic Letter 81-07, Attachment (3), required the following assumptions be provided: 1. UO / Water V lume Ratio - 0.61 2 2. Fuel Enrichment - Core average of the first cycle is 2.43% wt of U-235 3. TechSpecsrequireaK,gb2270 ppm,whicheverismorerestrictive. equal to or less than 0.95 or a boron concentration of at leas Comparison With Criterion III Criterion III - Damage to the reactor vessel or the spent fuel 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 horated source of adequate concentration if the water being lost is borated). Analysis - Three loads with the potential for damaging the reactor vessel are considered, the reactor vessel head, the plenum assembly and the missile shield segments. The head and the plenum assembly are lifted over an open reactor vessel while the missile shield segments are lifted over a reactor vessel with head installed. Before moving the plenum assembly the refueling canal is flooded. The plenum assembly must be raised 17 feet to clear the reactor vessel and indexing fixture before being moved laterally. A load drop of the plenum assembly would be through water (except for initial fuel load) which will slow the drop and substantially decrease the impact energy transmitted to the core barrel and reactor vessel. In order for the plenum assembly (55 tons) to fall and impact the core support assembly, after it has been lifted clear of the reactor vessel, the lifting rig would have to fail completely. The lift rig used for the plenum assembly is designed to lift the combined weight of the core support assemblies (180 tons) and therefore complete failure when lifting the plenum assembly is extremely unlikely. x s
19 f3 O There are four missile shield segments each weighing 52.5 tons, that span the refueling canal during normal operation. Prior to removal of the head for refueling each shield segment is moved to its storage location spanning the "D" ring. The shields are located approximately 26 feet above the top of the reactor vessel head. To preclude the potential for a missile shield drop onto the reactor vessel head, administrative controls allow each shield segment to be lifted vertically no more than three feet, moved laterally east or west to clear the reactor vessel head and then moved to its stowage location spanning the "D" ring (see Figure C-16). A plenum assembly load drop would be buffered by water (except for initial fuel load) in the refueling canal and reactor vessel and strict administrative controls will prohibit the free lif t of a missile shield segment over the reactor vessel. Since the reactor vessel head is three times the weight of the next heaviest load lifted over the reactor vessel, CP Co considered this the bounding load and elected to perform a head drop analysis. B&W performed a two phase head drop analysis: one phase considered a uniform load drop; the second phase considered an oblique (point) load drop. Results of these analyses indicated that a head drop of no more than 10 feet for the uniform drop case and a head drop of no more than 8'-8" for the oblique drop case did not violate NUREG-0612, Criterion III (p) requirements. The maximum height the reactor vessel head may be lifted \\~ / above the reactor vessel will be administratively controlled at 8.5 feet. Results of the head drop analyses indicated displacements of the core flood lines as follows (the core flood lines are of a concern since the DHR supply uses a portion of these lines for injection into the RCS): Vertical (Downward) Radial Displacement (Inches) Displacement (Inches) Uniform Drop .509 .033 Oblique (Point) Drop .336 .477 Supplementary analysis of the core flood lines is being performed. Results will be submitted as soon as they become available. Further information on the reactor vessel head drop analysis, including the information requested by Appendix A of NUREG-0612, is included in Appendix B of this report. /~'T t w./ -c.
20 k 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 Item 1 (2.4.1) Identify any cranes listed in 2.1.1 (Part I Response), 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 (ie, 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-I handling-system (ie, crane-load combination) information specified in. e l Response - Handling systems in the reactor building do not take credit for single failure proof criteria for all loads to be carried. See 2.4.2 below for the reactor building load handling evaluation. 1 (s
21 v Item 2 (2.4.2) For any cranes identified in 2.1.1 (Part I Response), not designated as single-failure proof in 2.4.1, a comprehensive hazard evaluation should be provided which includes the following information: Item 2a (2.4.2a) 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 include designation and weight or cross-reference to information provided in 2.1.3c (Part I Response). Impact areas should be identified by 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 - Load / Safe Shutdown Area Matrix, Tables 1 and 2, list all heavy loads, in the reactor and auxiliary buildings respectively, and potential impact areas where damage might occur to safety-related equipment. These loads are taken from Tables 3 and 4 of CP Co's Part I. Tables 1 and 2 reference load path figures and list potentially impacted equipment by elevation. Drawings of the load paths are included in this report, see Figures C-1 through C-18. Figures C-9, C-14 and C-16 have been revised f-s for Part II. (,, 4 O l l +
22 ('~N e 1 \\s_/ Item 2b (2.4.2b) 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 electri:al interlocks, or other site-specific considerations. Elimination on the basis of the aforementioned considerations should be suppler.ented by the following specific information: 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 (ie, the ability of the system to perform its safety-related function). 2. Where mechanical stops or electrical interlocks are to be provided, present details 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 removeable stops, for verifing 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. m I \\ s _,/ 3. Where load / target combinations are eliminated on the basis of other site-specific considerations (eg, maintenance sequencing), provide present and/or proposed technical specifications and diccuss administrative procedures or physical constraints invoked to ensure the continued validity of such considerations. Response - 1. Load / Safe Shutdown Area Matrix, Tables 1 and 2, include a hazard elimination category column. Where credit is taken for separation and redundancy the load path for a particular load is designed to ensure that only one of the redundant trains can be affected by a load drop. Utilizing separation and redundancy as a hazard elimination catagory in Tables 1 and 2 will require advanced maintenance planning to ensure that when a load is being carried over one system train the redundant train is operable. 2. There are no electrical interlocks or mechanical stops provided that preclude crane travel over safe shutdown equipment. 3. Maintenance schedules will include input of the heavy load lifts to be made and will sequence work such that equipment redundancy will not be compromised. Maintenance procedures will require that no heavy load lif ts be made over redundant equipment necessary for obtaining or maintaining safe shutdown conditions at the time the gg lift is made. This will be a requirement only if the heavy load safe (v/ load path is based on system redundancy.
23 I Item 2c (2.4.2c) For interactions not eliminated by the analysis of 2.4.2b 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 (ie, 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 (ie, crane-load combination) information specified in Attachment 1. Response - As noted in 2.2.3 above, the only handling system that will make the likelihood of a load drop extremely small through compliance with NUREG-0612, Section 5.1.6, is the one for the spent fuel cask (Table 2, Item 7). information is included in 2.2.3 response. ) l l l ) i l l
24 \\ Item 2d (2.4.2d) For interactions not eliminated in 2.4.2b or 2.4.2c, 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). 2. The basis for any exceptions taken to the analytical guidelines of NUREG-0612, Appendix A. 3. The information requested in Attachment 4. Response - 1. Midland has one overhead handling system that is designed to retain a load during a seismic event, the main hoist of the auxiliary building crane (see Response to 2.2.3 above). The seismic calculations for the reactor buildir.g crane were based on an unloaded crane at rest during reactor operation. The trolley was considered to be located anywhere along the girder span along with the following design criteria: 11 psi /see maximum change in girder differential pressure rate a. loads due to LOCA. b. Seismic loads in accordance with acceleration response spectrum Curves. c. Combined LOCA and seismic loads. 2. CP Co takes exception to General Consideration 1 (10) of Appendix A of NUREG-0612 which reads " Credit may not be taken for equipment to operate that may mitigate the effects of the load drop if the equipment is not required to be operable by the technical specification when the load could be dropped." CP Co will utilize maintenance planning, sequencing and procedures to ensure that equipment redundancy is not compromised (see response to Item 2.4.2b above). 3. In containment a potential area of concern is loads passing through the equipment batch. The DHR return line from the reactor vessel to the pump suction and one of two DHR supply lines pass directly under the equipment hatch. Since the return piping is a single lir.e for ) much of the run, it is susceptible to a heavy load drop. The DHR
25 O supply piping to the reactor vessel is redundant and sufficient physical separation is provided to preclude a complete loss of DHR injection due to a heavy load drop. A solution to this concern exists and is documented in Note 7 of Table 1. CP Co will provide an emergency procedure detailing the actions required in the event that the DHR return line integrity is compromised by a heavy load drop. Potential areas of concern in the auxiliary building include: a. Componept Cooling Water (CCW) Train Location - CCW Train A is i located one level above and directly over CCW Train B. The CCW system provides cooling water to selected nuclear auxiliary components during normal plant operation and to engineered safety features systems during a loss-of-coolant accident. The CCW heat exchangers are located on elevation 599'-0" for Train A and elevation 584'-0" for Train B, between north / south coordinates B & C and East / West coordinates 5.1 through 7.9 (see Part I Response, Figures A-1 and A-2). b. Makeup Pump Suction Lines - in one area of the auxiliary building the various suction lines to the makeup pumps are in close proximity to one another. The makeup pumps can take suction from the borated water storage tank or the makeup tank. If both suction lines are out of service no means of high pressure boron injection into the reactor coolant system would be available. The makeup pump suction lines are located on elevation 599'-0" between north / south coordinates D and E and east / west coordinates 5.6 through 5.9 and 7.1 through 7.4 (see Part I Response, Figure A-2). None of the safe load paths go over this area with the exception of the path for the demineralizer plugs (11.5 tons), Item 24 of Table 2. These plags are located directly over the makeup suction piping. CP Co does not foresee any maintenance that requires lifting these plugs and therefore considers this lift as an extraordinary maintenance item to be handled on a case basis. c. Auxiliary Building Electric Monorail - Loads handled with the electrical monorail include the main steam valve and process steam valve operators, Items 15 and 16 of Table 2. Analyses are i underway to determine the effects ' a load drop onto steam piping and onto the roof of the auxxiiary building. The results of these analyses will be documented and submitted as a supplement to this report. d. Transfer Tube Access Shield Plates - Item 31 of Table 2, the Shield Plates weigh 1.05 tons and are located over safe shutdown equipment piping. CP Co does not foresee any maintenance that i requires lifting this load and therefore considers this lift as an extraordinary maintenance item to be handled on a case basis. Because of the CCW and makeup pump suction line concerns a load drop n%, analysis was performed for the refueling level -(El 659'-0") floor of the auxiliary building. The results of he analysis shows that the i . - _ _, _ ~.
26 floor can withstand a load drop of up to 6 tons from 6 inches if the drop is on one of the structural members that run north and south under the floor at 8 foot centers. Typical paths over the CCW heat exchangers show the load being lifted along the structural floor supports. Further information on this analysis, including the information requested in Generic Letter 81-07, Attachment 4, is contained in Appendix C of this report. l Most of the loads listed in Table 2 will be bounded by the 6 tons 6 l inch limitation. Loads in excess of 6 tons or loads that must be lifted more than 6 inches will be transported across sensitive areas on an equipment carriage. Maintenance sequencing will be used to take into account heavy loads being lifted over redundant trains (see Response to 2.4.2b above). I k i i ( l
r r i i* TABLE I LOAD / SAFE SIIUTDOWN AREA MATRIX FOR REACTOR BUILDING ( llaza rd (2) Eliminat{gg Load Path Item Load (tons) Figure Safe Shutdown Equipment Within Lift Area Category i Reactor Vessel llead (165) C-6 El 634'-0" Reactor vessel (IT-51, 2T-51) A( } A(5) 4) 2 Plenum Assembly (55) C-6 El 634'-0" Reactor vessel (IT-51, 2T-51) B( El 587'-11" Reactor building sump 3 Inservice Inspection Tool (4.5) C-8, C-16 El 627'-0" Reactor coolant pressure B(6) boundary (reactor coolant pumps, steam generators, pressurizer, piping) El 593'-6" Decay heat removal system B inlet and outlet to reactor coolant system 4 Reactor Coolant Pump Hotor (50) C-16 El 627'-0" Reactor coolant pressure B boundary (reactor coolant pumps, steam generators, Pressurizer, piping) El 593'-6" Decay heat removal system B( } inlet and outlet to reactor coolant system 5 Stud Tensioners (1) C-7 El 634'-0" Reactor vessel A(
Table 1-(Continued) 2 Hazard Load Path Ites Load (tons) Figure Safe Shutdown Equipment Within Lift Area (2) Eliminatg Category 6 Crane Load Block B-10 El 627'-0" Reactor coolant pressure B -Main Hook (3.4) boundary (reactor vessel, reactor coolant pumps, steam generators, pressurizer, piping) El 593'-6" Decay heat removal system B( inlet and outlet to reactor coolant system 7. Snubbers (1-6) C-5, C-16 El 627'-0" Reactor coolant pressure B( boundary (reactor coolant pumps, steam generators, pressurizer, piping) El 593'-6" Decay heat removal system B( inlet and outlet to reactor coolant system 8 Boom Crane (3.6) C-15, C-16 El 627'-0" Reactor coolant pressure B( boundary (reactor coolant pumps, steam generators, press v izer, piping) El 593'-6" Decay heat removal system B( )' inlet and outlet to reactor coolant system 9 Letdown Cooler (3) C-2, C-16 El 627'-0" Reactor coolant pressure B(0} boundary (reactor coolant pumps, steam generators, pressurizer, piping) N El 593'-6" Decay heat removal system B(7) ) inlet and outlet to reactor coolant system l l i
Table 1 (C t ined) 3 llazard 3 Load Path Eliminat l Item Load (tons) Figure Safe Shutdown Equ_ipment Within Lift Area ( Category 10 Seal Plate (6) C-5 El 634'-0" Reactor vessel (IT-51, 2T-51) A( ) 11 Missile Shield (52.5) C-16 El 627'-0" Reactor coolant pressure B(6) boundary (reactor vessel, reactor coolant pumps, steam generators, pressurizer, and piping) 12 E<1uipment Carriage C-5, C-7 El 593'-6" Decay heat removal inlet 'I } and outlet to reactor coolant system (9) 13 Plenum Assembly and Core C-6 El 634'-6" Reactor vessel (IT-51, 2T-51) Support Assembly (180) 14 Reactor Vessel Top lleat Insu-C-5, C-17 El 627'-0" Reactor coolant pressure boundary H lation Hack Assembly (3.8) (reactor vessels, reactor coolant pumps, steam generators, piping) I (0) 15 Internals Indexing Fixture C-9 El 634'-6" Reactor Vessel (IT-51, 2T*51) (6.5) 4 i
O Q J 4 ) Table 1 (Continued) NOTES: (I Dif ferences in reactor building layout for Units 1 and 2 have been considered in the evaluations of load drop. Reactor will either be in Mode 5, cold shutdown (average coolant temperature at or below 200*F), or Mode 6, refueling (average coolant temperature at or below 140*F) before using the polar crane. (2) Equipment necessary to mitigate the effects of a heavy load drop, allowing the unit to achieve or maintain cold shutdown. (3)llazard Elimination Categories A. Analysis demonstrates that crane failure ind load drop will not damage safe shutdown equipment. B. System redundancy and separation preclude loss of capability of system to perform its safety-related function following this load drop in this area. (4)See 3ppendix B ( Not required for safe shutdown with decay heat removal (DliR) system intact. (6) Load drop does not disrupt continued removal of decay heat. (7)DilR system piping is located below the equipment hatch at El 593'-6". The single DliR letdown line and one of the two DiiR return lines run in this area. A load drop accident just inside the reactor building (RB) equipment hatch may af fect these lines. If this load drop accident occurs, a coolant path can be established by taking suction for the DilR system from the HB emergency sump and returning the coolant through the intact DHR return line. If this load drop accident occurs during refueling, and the HB emergency sump cannot be used for DHR suction, an alternate flowpath can be established by taking suction from the spent fuel pool. A flowpath is then established from the spent fuel pool, through the DilR system, into the reactor vessel using the intact DilR return line and through the fuel transfer canal back to the spent fuel pool. If' the load drop accident. occurs during cold shutdown but with the reactor vessel head in place, DliR suction can be from the BWST initially and then from the RB sunp. l (8)Not procured at this time, therefore, weight is unknown. (9) Reactor vessel defueled before load lif t. i i t
4 O O TABLE 2 LOAD / SAFE. SHUTDOWN AREA MATRIX FOR AUXILIARY BUILDING Hazard Load Path Item Load (tons) Figure Safe Shutdown Equipment Within Lif t Area (3) Eliminat{gg Category i 1 Inservice Inspection Tool C-7, C-13 El 599'-0" Component cooling water heat A (4.5) exchangers (IE-73A, 2E-73A) Makeup /high pressure injection B(3) pumps'(IP-58A, 2P-58A) Makeup pump suction piping from B(4) makeup tanks (IT-38, 2T-38) HVAC equipment supporting equip-B ment above Electrical cables supporting B equipment above El.584'-0" Component cooling water heat A exchangers (IE-73B, 2E-73B) Decay heat removal heat ex-B changers (IE-60A, IE-60B, 2E-60A, 2E-60B) HVAC equipment supporting equip-B ment above Electrical cables supporting B equipment above El 568'-0" Electrical cables supporting B Channel A decay heat removal pumps (IP-60A, 2P-60A) and supporting HVAC equipment
m() Table 2 (C'7itinued) ^ o 2 llaza rd i Load Path Item Load (tons) Figure Safe Shutdown Equipment Within Lift Area (j) Eliminatg Category 2 Reactor Coolant Pump Motor C-7, C-13 El 599'0" Component cooling water lieat (50) exchangers (IE-73A, 2E-73A) Makeup /high pressure injection B(3) pumps (IP-58A, 2P-58A) Makeup pump suction piping from B(4) makeup tanks (IT-38, 2T-38) ilVAC equipment supporting equip-B ment above Electric cables supporting B equipment above (5) El 584'-0" Component cooling water heat exchangers (IE-73B, 2E-73B) Decay heat removal heat ex-B i changers (IE-60A, IE-60B, 25-60A, 2E-60B) r HVAC equipment supporting equip-B ment above Electrical cables supporting B equipment above El 568'-0" Electric cable supporting B Channel A decay heat removal pumps (IP-60A, 2P-60A) and supporting IIVAC equipment 3 Snubbers (1-6) C-7, C-13 El 599'-0" Component cooling water heat A exchangers (IE-73A, 2E-73A) 3) Makeup /high pressure injection B pumps (IP-58A, 2P-58A) g) Makeup pump suction piping from B makeup tanks (IT-38, 2T-38) HVAC equipment supporting equip-B ment above Electrical cables supporting B equipment above
~__... - O O Table 2-(Concinued) 3 llaza rd Load Path Item Load (tons) Figure Safe Shutdown Equipment Within Lift Area (3) Eliminat{gg Category El 584'-0" Component cooling water heat A exchangers (IE-73B, 2E-738) Decay heat removal heat ex-B changers (IE-60A, IE-60B, i 2E-60A, 2E-60B) 1[VAC equipment supporting equip-B I ment above Electrical cables supporting B equipment above El 568'-0" Electrical cables supporting B Channel A decay heat removal pumps (IP-60A, 2P-60A) and supporting ilVAC equipment 4 Boom Crane (3.6) C-15 El 614'-0" Makeup tank (IT-38) B(6) El 599'-0" Component cooling water heat A exchanger (IE-73A) Makeup /high pressure injection B(3) pumps (IP-58A, 2P-58A) Makeup pump suction piping from B(4) makeup tanks (IT-38, 2T-38) IIVAC equipment supporting makeup B pumps IIVAC equipment supporting com-B ponent cooling water equipment (Unit 1 only) Electric cables supporting B equipment above l k ,,.y 7.,--m----% g--- r.-v-----= e - - - - - - - - - - " ' ~ - - -
O O T_able 2 (Continued) 4 Hazard gg) Eliminat{g9 Load Path Itea Load (tons) Figure Safe Shutdown Equipment Within Lift Area Catego ry El 584'-0" Component cooling water heat A exchanger (IE-73B) Decay heat removal heat ex-B changers (IE-60A, IE-608, 2E-60A, 2E-60B) HVAC equipment supporting B equipment above Electric cables supporting B equipment above El 568'-0" Electric cables supporting B Channel A decay heat removal pump (IP-60A) and supporting HVAC equipment 5 Letdown Cooler C-7, C-13 El 599'-0" Component cooling water heat A (3) exchangers (IE-73A, 2E-73A) Makeup /high pressure injection B(3) pumps (IP-58A, 2P-58A) 4) Makeup pump suction piping from B makeup tanks (IT-38, 2T-38) HVAC equipment supporting equip-B i ment above l Electrical cables supporting B 1 equipment above El 584'-0" Component cooling water heat A exchangers (IE-73B, 2E-738) Decay heat removal heat ex-B changers (IE-60A, IE-60B, 2E-60A, 2E-60B) HVAC equipment supporting equip-B ment above Electrical cables supporting B equipment above
.O O O Table 2 (Continued) 5 llazard Load Path 3) Eliminat{gg Ites Load (tons) Figure Safe Shutdown Equipment Within Lift Area Category El 568'-0" Electrical cables supporting B Channel A decay heat removal pumps (IP-60A, 2P-60A) and supporting HVAC equipment ( 6 Equipment Carriage C-7, C-13 El 599'-0" Component cooling water heat exchangers (IE-73A, 2E-73A) Makeup /high pressure injection pumps (IP-58A, 2P-58A) Makeup pump suction piping from makeup tanks (IT-38, 2T-38) IIVAC equipment supporting equip-ment above Electrical cables supporting equipment above ( El 584'-0" Component cooling water heat exchangers (IE-73B, 2E-73B) Decay heat removal heat ex-changers (IE-60A, IE-60B, 2E-60A, 2E-60B) llVAC equipment supporting equip-ment above Electrical cables supporting equipment above El 568'-0" Electrical cables supporting Channel A decay heat removal pumps (IP-60A, 2P-60A) and supporting IIVAC equipment i 7 Spent Fuel Shipping Cask C-9, C-13 El 599'-0" Component cooling water heat C (15-110) exchangers (IE-73A, IE-73B) Electric cables supporting C makeup /high pressure injection pumps (IP-58B, 2P-58B) i i
O O V U Table 2 (Continued) 6 Ilaza rd Load Path Item Load (tons) Figure Safe Shutdown Equipment Within Lift Area (g) EHainat{g Category El 584'-0" Component cooling water heat C exchangers (IE-73B, 2E-73B) Electric cables supporting C equipment above and supporting IIVAC equipment Electric cables supporting make-C up/high pressure injection pumps (1P-588, 2P-58B) (10) 8 Neutron Source Shipping Cask C-9, C-13 El 599'-0" Component cooling water heat (12) exchangers (IE-73A, 1E-73B) (10) Electric cables supporting makeup /high pressure injection pumps (IP-58B, 2P-58B) ( 0) El 584'-0" Component cooling water heat exchangers (IE-73B, 2E-73B) (10) Electric cables supporting equipment above and supporting HVAC equipment (10) Electric cables supporting make-up/high pressure injection pumps (1P-58B, 2P-58B) 9 Irradiated Specimen Shipping C-9, C-13 El 599'-0" Component cooling water heat A( } Cask (3.5-12) exchangers (IE-73A, IE-73B) Electric cables supporting B(10) makeup /high pressure injection pumps (IP-58B, 2P-58B) 4
Table 2 (Continued) 7 liaza rd Load Path Eliminat Category {gg 3) Item Load (tons) Figure Safe Shutdown Equipment Within Lif t Area i-El 584'-0" Component cooling water heat A( i exchangers (IE-73B, 2E-73B) Electric cables supporting B(10) equipment above and supporting liVAC equipment Electric cables supporting make-B(10) up/high pressure injection pumps l (IP-58B, 2P-588) 10 New Fuel Shipping Container-C-II, C-13 El 614'-0" Hakeup tank (IT-38, 2T-38) B including 2 fuel assemblies (3-4) El 599'-0" Component cooling water heat A [ exchangers (IE-73A, 2E-73A) f Makeup /high pressure injection B(3) i pumps (IP-58A, 2P-58A) Makeup pump suction piping from B(4) makeup tanks (IT-38, 2T-38) HVAC equipment supporting equip-B i ment above i I Electric cables supporting make-B(8) up/high pressure injection pumps (IP-58B, 2P-58B) t i 4 i t i I 1
,~ l N_ '1 %j i T:ble 2 (Continued) 8 Ilazard Load Path Item Load (tons) Figure Safe Shutdown Equipment Within Lift Ar,ea(3) Eliminat{gg Category El 584'-0" Component Cooling Water llea; A Exchangers (IE-73B, 2E-73P) Component cooling water pimp B (OP-73) Decay heat removal heat ex-B changers (IE-60A, IE-60B, 2E-60A, 2E-60 IIVAC equipment supporting equip-B ment above Electric cables supporting make-B(8) up/high pressure injection pumps (IP-58B, 2P-58B) 11 Fuel Transfer Carriage (2.5) C-10, C-13 El 599'-0" Component cooling water heat A exchanger (IE-73A, 2E-73A) Makeup /high pressure injection B(9) pumps (IP-58C, 2P-58C) Electric cables supporting make-B(9) up/high pressure injection pumps (IP-58B, 2P-58B) El 584'-0" Component cooling water heat A exchangers (IE-73B, 2E-73B) Electric cables supporting equip-B ment above and supporting IIVAC equipment Electric cables supporting make-B up/high pressure injection pumps (IP-58B, 2P-58B) Decay heat removal System B B piping (both units) El 568'-0" -Decay heat removal System B B piping (both units)
Table 2 (Continued) 9 Ilazard Load Path Item Load (tons) Figure Safe Shutdown Equipment Within Lift Area (g) Eliminatg Category l 12 Crane Load Block-Hain llook B-7, B-8 El 614'-0" Makeup tank (IT-38, 2T-38) C l (5.8) l i El 599'-0" Component cooling water heat C l exchangers (IE-73A, 2E-73A) i Component cooling water pumps C (IP-73A, 2P-73A) Makeup /high pressure injection C pumps (IP-58A, IP-58B, IP-58C, 2P-58A, 2P-58B, 2P-58C) IIVAC equipment supporting equip-C ment above Electrical cables supporting C equipment above l l l El 584'-0" Component cooling water heat C l exchangers (IE-73B, 2E-73B) l Component cooling water pumps C (OP-73, IP-73B, 2P-73B) Decay heat removal heat ex-C changers (IE-60A, IE-60B, 2E-60A, 2E-60B) IIVAC equipment supporting C l equipment above Electrical cables supporting C equipment above Electrical cables supporting C makeup /high pressure inj ection pumps (IP-58B, 2P-588) Decay heat removal system B C piping (both units).
s s, \\ \\ Table 2 (Continued) 10 Hazard Load Path Item Load (tons) Figure Safe Shutdown Equipment Within Lift Area (3) Eliminat{gg Category El 568'-0" Decay heat removal pumps C (IP-60A, 2P-60A) IIVAC equipment supporting C equipment above Electrical cables supporting C equipment above Decay heat removal System B C piping (both units) 13 Hakeup Pumps-3 per unit (3) C-2 to El 599'-0" Component cooling water heat A C-6, C-13 exchangers (IE-73A, 2E-73A) El 584'-0" Component cooling water heat A exchangers (IE-73B, 2E-73B) Decay heat removal heat B cxchangers (IE-60A, IE-608, 2E-60A, 2E-60B) Decay heat removal System B B piping (both units) El 568'-0" Decay heat removal pumps B (IP-60A, 2P-60A) ilVAC equipment supporting B equipment above Electrical cables supporting B equipment above Decay heat removal System B B piping (both units) 14 Auxiliary Feedwater Pumps C-1 to El 599'-0" Component cooling water heat A -2 per unit (2.8) C-6, C-13 exchangers (IE-73A, 2E-73A) El 584'-0" Component cooling water heat A exchangers (IE-73B, 2E-73B) 4 b
Table 2 (C5dtinued) 11 llaza rd Load Path 3) Eliminat{gg Item Load (tons) Figure Safe Shutdown Equipment Within Lift Area Category l El 568'-0" Decay heat removal pumps B (IP-60B, 2P-60B) Electrical cables supporting B equipment above 15 Main Steam Isolation Valve C-8, C-18 El 704'-0" Main steam isolation valves (Later) Operators (7.5) (both units) Main steam line piping (both units) El 685'-0" Control panels for auxiliary building flVAC and Unit I reactor building ilVAC El 659'-0" Safety equipment room El 646'-0" Electrical cables supporting engineered safety features t actuation system El 614'-0" Emergency batteries (both units) Electrical cables supporting Channel A safe shutdown equip-ment (both units) Safety related electrical switch-gear room El 599'-0" Electrical cables supporting makeup /high pressure injection j pumps (both units) i f l
Table 2 (Continued) 12 llaza rd item Load (tons) Figure Safe Shutdown Equipment Within Lift Area (3) Eliminat{gg Load Path Category El 584'-0" Auxiliary feedwater pumps (IP-05B, 2P-05B) Electrical cables supporting makeup, component cooling water and auxiliary feedwater (both units El 568'-0" Electrical cables supporting Channel B decay heat removal (both units) 16 Process Steam Transfer Valve C-8, C-18 El 704'-0" flain steam isolation valves (Later) Operators (7) (Unit 2) Hain steam line piping (both units) 1 El 685'-0" Control panels for auxiliary l building IIVAC and Unit I reactor building ilVAC l l El 674'-6" Upper cable spreading room El 659'-0" Main control room and safety equipment room El 646'-0" Lower cable spreading room Electrical cables supporting engineered safety features actuation system El 614'-0" Safety-related electrical switchgear room Emergency batteries (Unit 2) Electrical cables supporting Channel A safe shutdown equip-ment (both units)
O O O . Table 2 (Conranued) 13 Hazard Load Path Item Load (tons) Figure Safe Shutdown Equipment Within Lift Area (3) Eliminatg Category El 599'-0" Electrical c 51eu supporting makeup /high pressure injection pumps (Unit 2) El 584'-0" Auxiliary feedwater pump (2P-05B) Electrical cables supporting makeup, component cooling water, and auxiliary feedwater (both units) El 568'-0" Electrical cables supporting Channel B decay heat removal (Unit 2) 17 Filter llandling Machine (5.5) C-12, C-13 None 18 Filter Transfer Cask (2.6) C-12, C-13 None 19 Equipment Access liatch - C-1 to El 614'-0" Makeup Tank (IT-38, 2T-38) B 6 places (1.25) C-5, C-14 El 599'-0" Component cooling water heat A exchangers (IE-73A, 2E-73A) Component cooling water pumps A (IP-73A, 2P-73A) Makeup /high pressure injection B(3) pumps (IP-58A, 2P-58A) Makeup pump suction piping from A makeup tanks (IT-38, 2T-38) and BWST HVAC equipment supporting equipment B above Electrical cables supporting make-B(8) ) up/high pressure injection pumps (IP-58B, 2P-588) 4* 1 1
3 O Table 2 (Continued) 14 1 liaza rd j) Eliminat{gg Load Path Ites Load (tons) Figure Safe Shutdown Equipment Within Lift Area Category .El 584'-0" Component cooling water heat A exchangers (IE-73B, 2E-738) Component coolang water pumps A (OP-73, IP-73B, 2P-73B) Decay heat removal heat B exchangers (IE-60A, IE-60B, 2E-60A, 2E-608) IIVAC equipment supporting B equipment above Electrical cables supporting make-B(8) up/high pressure injection pumps (1P-58B, 2P-58B) 20 Decontamination Area Access C-15 None flatch, Plug No 1 - 1 place, 2 pieces (6.9) 21 Filter Plug No 2 - 24 places C-14 None -(6.8) 22 Spent Resin Access llatch, C-15 None Plug No 3 - 1 place, 2 pieces (24) 4 23 Spent Resin Access Hatch, C-15 None Plug No 4 - 1 place, 2 pieces (20) 12 4 Demineralizer Plug No 5-C-14 El 614'-0" Makeup Tank (IT-38, 2T-38) B l 13 places (11.5) i i r
O J J Table 2-(Continued) 15 llazard Load Path item Load (tons) Figure Safe Shutdown Equipment Within Lift Area (3) Eliminatg Category (3) El 599'-0" Makeup /high pressure injection pumps (IP-58A, 2P-58A) (3 g Makeup pinnp suction piping from makeup tanks (IT-38, 2T-38) and BWST (33) HVAC equipment supporting equipment above i Electrical cables supporting (11) makeup /high pressure injection pumps (IP-588, 2P-58B) (II) El 584'-0" Decay heat removal heat exchangers (IE-60A, IE-60B, 2E-60A, 2E-608) gg IIVAC equipment supporting equipment above Electrical cables supporting make-(11) up/high pressure injection pumps (IP-58B, 2P-58B) 25 Degasifier Plug No 6-C-14 El 614'-0" Makeup Tank (IT-38, 2T-38) B 2 places (11.7) (II} El 599-0" Makeup /high pressure injection pumps (1P-58A, 2P-58A) gg Makeup pump suction piping from makeup tanks (IT-38, 2T-38) and BWST IIVAC equipment supporting equipment above gg Electrical cables supporting make-i up/high pressure injection piumps (IP-58B, 2P-58B)
i ~- Table 2 (Continued) 16 Ilaza rd Load Path item Load (tons) Figure Safe Shutdown Equipment Within Lift Area (3) Eliminat{gg Category (I El 584'-0" Decay heat removal heat exchangers (IE-60A, IE-60B, 2E-60A, 2E-60B) (33) IIVAC equipment supporting l equipment above (33) Electrical cables supporting make-up/high pressure injection pumps (IP-58B, 2P-588) 26 Pipe Chase Access Plug No 7-C-15 El 614'-0" Electrical cables supporting B 2 places (2.2) Channel A component cooling water, makeup pumps and decay heat removal system (Unit 2 only) El 599'-0" Electrical cables supporting B Channel A component cooling water, makeup pumpa and decay heat remova' system (Unit 2 only) El 584'-0" Electrical cables supporting B Channel A decay heat removal system (Unit 2 only) 27 Pipe Chase Access Plug No 8 C-15 El 614'-0" Electrical cables supporting B (1.9) Channel A component cooling water, makeup pumps and decay heat removal system (Unit 2 only) El 599'-0" Electrical cables supporting B Channel A component cooling water, makeup pumps and decay heat removal system (Unit 2 only)
t,,, (n ) Table 2 (Continued) 17 Ilazard Load Path Iten LosJ (tons) Figure Safe Shutdown Equipment Within Lift Area (g) Eliminat{gg Category El 584'-0" Electrical cables supporting B Channel A decay heat removal system (Unit 2 only) 28 Pipe Floor Chase Shield C-4 El 599'-0" Electrical cables supporting B Plug No 9 - 2 places (3.9) auxiliary feedwater and component cooling water 29 Valve Pit Steel Deck - C-1 None 2 Places (1.1) 30 Valve Pit Steel Deck - C-1 None 2 places (1.3) ( ) 31 Transfer Tube Access, C-4 El 568'-0" Decay heat removal piping 3 Shield Plates (1.05) Auxiliary feedwater piping Hakeup/high pressure injection P ping i Electrical cables for high pressure injection and auxiliary feedwater valves NOTES: (1) Equipment necessary to mitigate the effects of a heavy load drop, allowing the plant to achieve or maintain cold shutdown. (2)llazard Elimination Categories A. Analysis demonstrates that crane failure and load drop will not damage safety-related equipment. B. System redundancy and separation precludes loss of capability of system to perform its safety-relatcJ function following this load drop in this area. C. Likelihood of handling system failure for this load is extremely small.
.. ~ p-~ s k, s. 18 (3) Redundant makeup pumps, including supporting IIVAC equipment and cables, outside lift area. (4) Common suction piping for all makeup pumps from makeup tank in load path. Piping and valves for alternate makeup pump suction source (borated water storage tank) outside lift area. (5) Load will not be lifted over listed equipment. (6) Piping and valves for alternate makeup pump suction source (borated water storage tank) ou*, side lift area. (7)Not procured at this time, therefore, weight is unknown. (8) Single load drop cannot affect electrical cables supporting both Channel A and Channel B makeup pumps. t (9) Single load drop cannot. affect both Channel C makeup pump and electric cables supporting Channel B makeup pump. (10)If this lif t is' over 6 tons, consideration will be given to transport to the railroad bay on a carriage (ie, no lift except on and off the carriage.) (11) Plug removal necessary to gain access to demineralizer/degasifier tanks. Access to tanks is considered extraordinary maintenance and will be addressed on a case-by-case basis. (12) Shield plate removal necessary to gain access to external surface of fuel transfer tube. Access to external surface of i fuel transfer tube is considered extraordinary maintenance and will be addressed on a case-by-case basis. 1 l j l s y
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== !E} LIFT AREA WITHIN LOAD PATH LOAD PATHS n :=t, AT ELEVATION 614'-0" "U SAFE SHUTDOWN EQUIPMENT FIGURE C-3 d AND FUEL STORAGE AREAS G-1848-23
i ( Q Q y.11C Q('4 .- = _u., w_ E PA1 SEE FIG C-13+1;,vpM:__ SO C-1 gp ~in , _. _. ~.. W9.E..TIF=yza.M,#c-i.=:ce_=.= g....... i Lbm. w'm i + U b. 3 M e . e. - 2 / ~ g PATH 6F SHIELD PLUG Nd. 9 9f:3b ~bs.";..: - G U Wg'.%[_ ]#;3.p,gK@%,f.;.,.]yt {; e, r. c.r; i'_2gh_Z b y ~tc ~ r.,. / N +,
- c.. -
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- 1. I~ 2 L
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- =4A.AS*
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- r s
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g g,,,gg,,g, y I. J% ' MI C I b$ e6 4 Gace.g's 3v3 stanctS (a6.h4'
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\\ \\ F1.9 v' t ..az=e,== u-b..,l-w.gg-,3Eg I, T ^~ - r-1 "H OF FILTER TRANSFER CASK TO ') t t .b ~ LID RADWASTE BUILDING, SEE FIG mg x t r r t r g y/J {4 t g, ..a;p o 5 e + e k(.$Y,. o $? !$0 ?hY -l EQUIPMENT HATCH LIFT TO g, N i j.1.. EL 659'-0", SEE FIG C-14 P' ~ ~ 3' f. ] '..r-w=EqE. a'A ! 1,e hd ut [ r -~ ,, a .- Oc ."a O ay ~ "5 3 s4.s]) f : ,( N. ... ','n u .>;..a.:,; ~. t r xy e glQ.,!D. g,gf:& q.{ATH OF SHIELD P A 'ix NDN 1 g I N .5 r1 g d.13. P -_r %~~m }l i~G Q '~ ~ ~~ t. Q ,N2=:, 2a - L%fw,.y fit g j .a .J1 g4/d -g ~--hn,f jt E' "O \\ c Q":.jf.'y}E ;! L T~ $$h.@ a
- 1i 1
-l - [3! p:t"2 B ,g y n PATH OF SHIELD PLATES 3 ,l, gr I / i L,L=u-O I um f ',,\\ R(-f-Q.. d / &:m)'M ' \\ y., u -3 i r \\ \\ x N;.N-e i. a-c u. , 'iQy (b* ( hL N m e T / w 4, Ir. ::=:-f Ek i WD 3 . pt.: nr p;ganza -4, 2 .. _11! w I r" ~ l L"*' M-"- - P j CONSUMERS POWER COMPANY 5===- : V !S LEGEND MIDLAND UNITS 1 AND 2 q ~~' LIMIT OF LOAD PATH HEAVY LOAD STUDY: NUREG-0612 LIFT AREA WITHIN LOAD PATH LOAD PATHS AT ELEVATION 634'-6" SAFE SHUTDOWN EQUIPMENT FIGURE C-4 bu AND FUEL STORAGE AREAS G-184ti-24 L
f r g g g - nr,! ._ _. _... _g a r - r- - - m.3 7.:.t g c s s,q. t.ur, w m. + i 4 /,) 7f:b ].-+' 4 w N/ p / '5 + /,, g -n A a a f, 'Y f **l* m i ****b 2' l& Y/ g p o.4 u. y.. w r.. % +'ry:gr, m i i i. "m 7 y n g.. . a +. :n-w w?r1.. v y.:u .- n. -- n ..... +.., n I ~- y _. ?p' M m'/r./ 7..
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PLATE (TYP) F.TO
- l
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e-- K' A,m.,,,,
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+,f f f i M^, w% $~ 4 2 m % jf.$._.:lq e..::~ d. ) l 7T %s e o e w t>fia!&:r-34,T/ir c4+F 9,.l. EQUIPMENT HATCH LIFT TO EL 659'-0", p SEE FIG C141 = ._.-4, ~.-
- n
.,r -..~ ;f y y ., m, m. y <c~~ .s C i $ '",7 [- g' M ~ k,1 i._ p_ ~'p 'i. E ='+4eHyn t-oCM1 > T '% ' PATH OF-MISCELLANEOUS EQUIPMENT (~- - y"
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'..r: _ o e I p ,...+ 2 Q..Q... ' Ml. ' a a.v - -y e. C .c g = =. g w $m;g ...: i. 2.:.- -, -g] ,,{ -F 10 Q89 M]_ _p - fN / t, .." PATH OF:8NUBBERS, ir.= - ?h'k f_Y l!&y 0 [$.f' w[.,M 4 W[ WWIF5'l 'g g g : f?. N IRTH OF REACTORtVESSEL HEADE 4 9dSULATION RACK ' SSEMBLY (TYP),: g,iW, #,.~.v A 9EE FIG!C-17 M "SIA V 41.4 p? Wo@W ,y;p, y gn .m-- p 4 b.vl 3@.yggs,.d V ' / r / b -." W - li., F -I : J-9 i i %; 4 g+4 (. -f1 ' ' Y i '!.' "_W' s E j i[, rs Ew r ag r@g.c =w Le_ - m# = < -- 9 ~_ ~. . as y >'e 4 1,.. q Ndo T ' g k CONSUMERS POWER COMPANY 7 ' "' ~ ~ ' I ~~ MIDLAND UNITS 1 AND 2 LEGEND A @L HEAVY LOAD STUDY: NUREG-0612 LIMIT OF LOAD PATH = = - LOAD PATHS LIFT AREA WITHIN LOAD PATH AT ELEVATION 645'-0" SAFE SHUTDOWN EQUIPMENT FIGURE C-5 l AND FUEL STORAGE AREAS ,4 G M'48-25 I
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- / <[' JASSEMBLYl(TYP AM V,
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ay I h 4-g g f g, 4l~~lW i 2 s<..- n ~.s 3deaw mene so &4P ..w, rga.ng bu.it 4A esvSecue. maat 978 0585 S Amh & g ='n.se *=oenres s=.e 'tipsb.e6 A k) g,o awoman n ,e r U
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- e.
~ 5 f5] 5] 5] % Q* MM' r o e t. + -g WG Ic
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4 g--- g g l9 s ( ' ~4 PATH IN! AUXILIARY BUILDING, , M c., (.7P SEE FIGEC-1, C-2 Up C ' 1 ~ M_M M [ d @ k i j.1; d k @3; r~:i y x,, g; 8 i-i O fR a. 1+
- n:.
Z p g .g g <g-a i, g.7 g i e r t...m__.M r. 7, _ J !, v,w o m. p a~ .. ~.. .a \\> sA /1W ?l g ~ -h ] c.,.. -- 4 PATH OF REACTOR - g k a._ - _4*. .H. EAD (TYP) my e------ i y p ny ,Cl,, ..... C 3} gg . 3.== g},4%s;s(j g.. 71 N gyy J mW 'iv
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~ f ih_. ' emu Fy, m GL ' S-E M. MM @d.- w Os. S g. LEGEND LIMIT OF LOAD PATH CONSUMERS POWER COMPANY , h LIFT AREA WITHIN LOAD PATH MIDLAND UNITS 1 AND 2 S SAFE SHUTDOWN EQUIPMENT HEAVY LOAD STUDY: NUREG-0612 AND FUEL STORAGE AREAS LOAD PATHS AT ELEVATION 659'-0" FIGURE C-6 f G-1DI8-26
( ( b f l w !:'.?L'**... ,$'y i j I L--..J (o. m j a::.y?,; "..,# 't Q M. ..._.i 1ll
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. _ g;r----IJTA!3 \\Tt: PATH TO R.R. BAY,d) h 'Sh S:,.-JW L_. N @&g-4 5% l j- - - i l x j [5}fdh U--- 1 L SEE Fl$(C-13 "- JQ N ]i "Q p ~:;*l'jd ".kl% M % MM-h; i & :fTi$e 1~.a--+Q.. M" :-&" ~F,,,} 4 T% H foi '#W'd-e i!o! ici 'of iot 5 $i,; a *e
- e
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- h..:..,. c ),a"= t**
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- =r-o 3 w, _N$33M_ 3k* 9[4
~ ' j)/&fay ,,,_...~.: 4 T li 8 4 C [l $ 4 L f D 1 j SEE FIG _C 16 A - d 7 N e a ~W I '., - A $h Ti. 5s A'%mMlW f t" 4 i (Q d'4 l f"WW 1%# 5ja i b db Mk--h 1 dg "/ - PATH OF HEAD STUD TENSIONERS. r FROM' EQUIPMENT' CARRIAGE (TYP),d.,. 7 4 Y =>SEE FIG-C-61 E. L J }i-g 'A.;. g-- g[gr ^OI5ICJ I. 7 u y
- /
1 "w. } n.. lM t.g A. -~ L/.w e (* " D. s '==. m pr a.or. 4-,, _..; aM LjANp'==i %.# ha-/ i. x n 1 i wc wm 1.g m b [.b d' 3 c
- MSMh8EbMNi E"'
s%g,i 6-W' '" M MT / 8 CONSUMERS POWER COMPANY LEGEND MIDLAND UNITS 1 AND 2 6+ LIMIT OF LOAD PATH HEAVY LOAD STUDY: NUREG-0612 E LIFT AREA WITHIN LOAD PATH LOAD PATHS AT ELEVATION 659'-0" ~ d/M SAFE SHUTDOWN EQUIPMENT . #dd AND FUEL STORAGE AREAS FIGURE C-7 G-1848 27/p
( O -~ Q z i t
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- iwr 1
u PARTIAL PL AN j_ 9 ELEVATION 67TT 4 QQ -Qg > b c % : rw . =% a t ;.-,.v;r
- n.
A % 5it&+ s -w 7 U l;-f ; ~ =- M Ghi$a n ~,w %. w.n ,d P. r:. s x q [ I,. s. ,...,y p _[ &l f -7 [. I\\ d I. .. [ J i C L p -( gy ..f 1 ) W -'hy4b /
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PATH OF ASSEM.. BLED INS-m Bgi 'y f T SEE FIG C-1_6M 1 M a INSPECTION TOOL (TYP) : w p ?
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== a.s.s g p - eP.e.t04bd d.m*WGer. = ,,,,,,,,.m d8" Tir.88 TW.8'E887 W C ep p y p,ange ge. mm --~ .-_.-,,~,.DEU,'"* . I. _.._.... est ..-u~-~~+.".'=.-.,--s'r 'C' WT N h4 1 ~ g c .f.>P,L,DA_as -c SS.A( C cas C68 gae .u.... &v
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- _9_ d9j [9.} [9.} } }
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,h by d."W qf m:,< Ci!F9 Tj, Q-fgjisliiafi 8... e... H M \\p p r m L.iy +[;,$, a-6'$. $.! M _..= - s -e = f8 Q F J -a. o 's - 3 r L Ql#$l 0 K C@ PATH OF, MADI STEAM ISOLATIO /lMitG)h. ~ hA !.- v ~ Al,,v d ___ r $ 5 %[ (" 'VALVEfAND PROCESS STEAM 0.7, __ odd , TRANSFER. VALVE OPERATORS e g' RVICE 4' n -5 h - ---,-~r bcAt bfQLh-MlLe' p. 7"=;- ..N G, s^ SEE FIG C-18 :- - - - - - - - )-- fg- )%K M s f ;/ bgr-... g.
- p., N--
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p
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a g,g + j.6 o w nm .I w M,.._zgE.,,1.@ g g bulm-lg l,g f 7,,.. e, e ..,.c,.\\ m., i.iiL S_ ~ ~ e s LEGEND CONSUMERS POWER COMPANY LIMIT OF LOAD PATH MIDLAND UNITS 1 AND 2 r-" HEAVY LOAD STUDY: NUREG-0612 LIFT AREA WITHIN LOAD PATH LOAD PATHS SAFE SHUTDOWN EQUlPMENT AND FUEL STORAGE AREAS AT ELEVATION 659'-0" FIGURE C8 anysde
h A' 6'_____~. 9g I'i i e-i men @ I -1*L*~, *... y I = c q [ $*;' ij r. .T in ,Oq I
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- 3-n.-
w. m zw %-ad.n tH. yo ,, - -~n y 4 wp ff, W
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is " y ((fj I I' I / s a, s n O [_ - t h4 4 same pu L. N q ,,,,4ghA ' M{' WWhfth ili"Ml.fMN ' ggp' PATH-OF IN _/ 1.E kh ihr j 8 4
- 5. " *......s...........
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/ [ S I LEGEND CONSUMERS POWER COMPANY LIMIT OF LOAD PATH MIDLAND UNITS 1 AND 2 { LIFT AREA WITHIN LOAD PATH LOAD PATHS SAFE SHUTDOWN EQUIPMENT AT ELEVATION 659'-0" AND FUEL STORAGE AREAS
- i FIGURE C-10 G-1843:*50
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1, i i: mem 2 1 - 'i* _W* git' Q;g. k WV= 3W MY hYh _ __g -r-- - __... ; s LEGEND CONSUMERS POWER COMPANY %~ LIMIT OF LOAD PATH MIDLAND UNITS 1 AND 2 = HEAVY LOAD STUDY: NUREG-0612 5__ m] LIFT AREA WITHIN LOAD PATH f_::;y: e. SAFE SHUTDOWN EQUIPMENT LOAD PATHS AND FUEL STORAGE AREAS AT ELEVATION 659'-0" FIGURE C-11 J G-18(8-31
( f ( PATH OF FILTER TRANSFER CASK TO h, SOLID RADWASTE BUILDING, 4
- . :e SEE FIG C-4
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19g yEj ~ , ta - 3 I m. _ < es mm 41 4 m .T ~ ~. . v. n et er a' k:I.- [.. M$N N '~~'. g 3. n g / j LEGEND CONSUMERS POWER COMPANY 9a LIMIT OF LOAD PATH MIDLAND UNITS 1 AND 2 P84.pe? HEAVY LOAD STUDY: NUREG-0612 LIFT AREA WITHIN LOAD PATH SAFE SHUTDOWN EQUIPMENT LOAD PATHS AND FUEL STORAGE AREAS AT ELEVATION 659'-0" ~ FIGURE C-12 e a G-1848 *$2
i i C. .O .f4 6 aJ
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- I.
N T l' ,I ~ - PATH CONTINUES INTO SOLID RAD- ) a i WASTE BUILDING ...e ,o ~ u e a.e-9 9. s.. n .i f.! f I l .) [ l 1.T=.'. ^j f ' ( n v n . fi y.,i. Q-,{ j -ymi#W4h- @ k b(l~ l @^ ' U,'M. ' J c 3, e' N_ g s- - - g._ r .; : ;m...... #
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- e ~
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- s. -
t'" / SECTCN GG '4 LOAD PATHS AT ELEVATIONS 659'-0" & 634'-6" FIGURE C-13 j/ G-1848-33
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O CONTROL OF HEAVY LOADS AT NUCLEAR POWER PLANTS i (NUREG-0612) PART II RESPONSE APPENDIX A SUPPLEMENTARY INFORMATION FOR OFF-SITE RADIOLOGICAL RELEASE ANALYSIS, REACTOR BUILDING O l l 7 rO-t
L NUREG-0612 { Part II Response ( Appendix A ) TABLE OF CONTENTS I s i ( o. ' Excerpts From Midland FSAR i Subsection / Table /Q&R +' s i '73.7.4-Fuel Handling Accident = 1 Table 15.7-5 Parameters Used in Evaluating the Radiological Consequences of a Fuel Handling Accident Table 15.7-7 Radiological Consequences of a Postulated Fuel Handling Accident Inside Containment i 12.3.4.2 Airborne Radioactivity Monitoring j Q&R 312.44 (15.7.4) 9.4.9 Reactor Building Air Purification and Cleanup System s' a I i h I 4 J w i \\ } 1 .E --~.m.__,,--.__2 .,-.,5,. .,,,_,,,,.,y,, -.~m.,.---
NUREG-0612 FART II RESPONSI l MIDLAND 1&2-FSAR APMDQIX A l g-~ 15.7.3 POSTULATED RADIOACTIVE RELEASES DUE TO LIQUID-CONTAINING l t TANK FAILURES l \\_/ The analysis of a liquid-containing tank failure resulting in the l I contamination of surface waters is presented in Subsection 2.4.12. A similar analysis resulting in the contamination of groundwater is presented in Subsection 2.4.13. 15.7.4 FUEL HANDLING ACCIDENT l 15.7.4.1 Identification of Causes and Frequency Classification i The possibility of a fuel handling accident, classified as an l infrequent incident, is remote because of the many administrative l controls and physical limitations imposed on the fuel handling l operations. All core alterations after the initial fuel loading chall be performed either by a licensed reactor operator under the general supervision of a licensed senior reactor operator or by a nonlicensed operator directly supervised by a licensed senior 20 reactor operator (or licensed senior operator limited to fuel handling) who has no other concurrent responsibilities during i this operation. l l When transferring irradiated fuel from the core to the spent fuel l pool for storage, the reactor cavity and refueling cavity are (Sfilled with borated water et a boron concentration equal to that
- l Jin the spent fuel pool, which ensures subcritical conditions in
\\the core even if all control rod assemblies are withdrawn. After the reactor head and control rod drive shafts are removed, fuel assemblies are lifted from the core, transferred ver*.ically to l the upender in the refueling canal which tips the fuel assembly onto the fuel transfer mechanism in a horizontal position. The transfer mechanism, which can accommodate two fuel assemblies, i transfers the fuel through the transfer tube to the tilt pit adjacent to the spent fuel pool where it is upended to a vertical position, transferred to the spent fuel pool, and then lowered into steel racks for storage in the spent fuel pool. l The fuel handling bridges are designed so that fuel cannot be raised above the minimum depth of water required for adequate chielding and protection of operating personnel. The containment, auxiliary building, refueling cavity, fuel transfer canal, and spent fuel pool are designed in accordance with Seismic Category I requirements, which prevent the etructures themselves from failing in the event of a safe chutdown earthquake. 15 All fuel handling equipment which operates above the fuel casemblies is designed to prevent it from f,lling and generating missiles which may damage the fuel. The construction of the (~' racks containing the fuel assemblies precludes damage to the fuel (_jin case portable or hand tools are dropped into the fuel pool. Revision 20 15.7-4 4/79
NURIC-0612 FART II RESPONSI MIDLAND 1&2-FSAR APPENDIX A The facility is designed so that heavy objects, such as the spent fuel shipping cask, cannot be carried over or tipped into the spent fuel pool. Movement of equipment handling the fuel is kept at low speeds. The design and use of interlocks prevent the fuel assembly from striking another object or structure during transfer from the core to its storage position. 15.7.4.2 Sequence of Events and Systems Ooeration The probability of a fuel handling accident is very low because of the safety features, administrative controls, and design characteristics of the facility previously discussed. However, it is postulated that a fuel assembly is dropped during 30 refueling operations in the aux:liary building or the containment, breaching the cladding of the fuel rods and releasing the volatile fission products contained in the gap region of the fuel rod. In the auxiliary building, a fuel assembly could be dropped in the fuel tilt pit, the spent fuel pool, or in the cask loading pit. In addition to the area radiation monitor located on the bridge over the spent fuel pool, portable radiation monitors capable of emitting audible alarms are located in this area during fuel handling operations. Doors in the auxiliary building are closed to maintain controlled leakage characteristics in the spent fuel pool region during refueling operations involving O spent fuel. Should a spent fuel assembly be dropped in the auxiliary building and release radioactivity above a prescribed level, the radiation monitors sound an alarm which alerts the personnel to the problem. In addition, the spent fuel pool ventilation actuation signal will be actuated on high airborne activity in the normal fuel pool exhaust header and shift the exhaust through the emergency fuel handling area exhaust filters to remove most of the halogens prior to discharge to the atmosphere. The operation of the spent fuel pool area ventilation system is further described in Subsection 9.4.2. In the containment building, a fuel assembly can be dropped during its movement from the reactor core to the fuel transfer tube. During fuel handling operations, the containment is kept in an isolable condition with all penetrations to the outside atmosphere, except for the equipment hatch, either closed or capable of being closed on an alarm signal from redundant radiation monitors. The reactor building purge air and refueling canal exhaust system is designed to provide adequate time to isolate the containment to minimize radioactivity released to the atmosphere as a result of a fuel handling accident. Redundant 30 Class lE radiation monitors are provided as part of this system to monitor the refueling canal area and system exhaust air for radiation and to initiate a signal to actuate RBIS-II upon detection of high radiation levels. As discussed in Subsection 9.4.9.2.3, a maximum of 1,430 ft 3 of containment air emanating from the refueling canal area would be released 15.7-5 Revision 23 4/81
EmIG-0612 MIDLAND 1&2-FSAR FART II RESPCNSE AP'ENDIX A (A through the purge air exhaust system penetration before the i ) isolatien valves are fully closed. In the calculation of this -d release, it is assumed that the activity is mixed with 10% of the containment volume prior to release. For further information on the operation of the purge air and refueling canal exhaust system, refer to Subsection 9.4.9. The equipment hatch, located at elevation 659'-0" (see Figure 1.2-8), will remain open during refueling operations. Although the equipment hatch represents a nonautomatically isolable penetration, it does not represent a direct release path to the environment because the hatch opens up to the new fuel and spent fuel pool storage areas in the auxiliary building. The spent fuel pool area is maintained at a slightly negative pressure by supplying air at 32,000 cfm and exhausting air at 36,000 cfm. The additional 4,000 cfm normally infiltrates through miscellaneous small openings throughout the auxiliary building. The net result of this design is that any radioactive release drawn through the equipment hatch would be exhausted 30 through the fuel pool ventilation system. Because the fuel pool ventilation actuation system (FPVAS) terminates normal supply and exnaust and actuates operation of the ESF grade standby exhaust system upon detection of high radiation in the exhaust air, the radiological consequence of leakage from a fuel handling accident in containment through the equipment hatch is limited to levels below that which have been analyzed for a fuel handling accident in the spent fuel pool area. In the calculation of a release 'sthrough the equipment hatch, it is conservatively assumed that the radiological consequence is the same as previously analyzed j for a fuel handling acc2 dent in the spent fuel pool area (see Table 15.7-6). For further information on the operation of the f uel pool ventilation system, refer to subsection 9.4.2. I The results of the analysis to calculate the total radiological consequences of a fuel handling accident in containment are presented in Table 15.7-7. 15.7.4.3 Core and System Performance This paragraph is not applicable for a fuel handling accident. 15.7.4.4 Barrier Performance The initial barrier limiting the transport of radioactivity from the damaged fuel to the environment following the postulated fuel handling accident is the water in the spent fuel pool or refueling pool. Decontamination factors of 100 for iodines and 1 for noble gases are used in the analysis. A second barrier between the released activity and the environment which applies only to a fuel handling accident in the auxiliary building is the emergency fuel handling area exhaust [,T O' r,, ) 15.7-6
- ev i:. t on 33 4/81 g
mm2C-0612 TART II RESPCNS MIDLAND 1&2-FSAR APMDmIX A gs filter system. Since the auxiliary building is designed Seismic Category I, the integrity of the auxiliary building will be preserved during the course of a fuel handling accident. This results in a controlled release of radioactivity through 4 inch deep bed carbon adsorber filters. Decontamination factors of 100 for iodines and 1 for noble gases are used in the analysis. These filters are described in Subsection 9.4.2. Barrier performance for a fuel handling accident inside of 3c containment is discussed in Subsection 15.7.4.2. 15.7.4.5 Radiological Consequences The major assumptions made for this analysis are shown in Table 15.7-5. The design basis analysis is consistent with the 13t assumptions in NRC Regulatory Guides 1.25 and 1.13. See Appendix 3A for a discussion of compliance with regulatory 32 guides. These regulatory guide assunptions are conservative in in the following respects: a. Radial Peaking Factor - The calculated value for the Midland units is 1.4. However, the value used in the analysis is 1.65 in accordance with the recommendations of NRC Regulatory Guide 1.25. 34 (~') b. The calculated iodine-131 gap activity is approximately (,f 2% of the fuel activity (refer to Table 15A-3). The gap activity used in the analysis is 10% of the fuel activity in accordance with the recommendations of NRC Regulatory Guide 1.25. In addition, no credit is taken for holdup of activity subsequent to its release from the damaged fuel assembly or for ground deposition or decay in transit to the boundaries of the exclusion area and low population zone. The design basis analysis assumes 3C damage to all 208 rods of a fuel assembly. The conservative analysis assumes damage to only the 56 peripheral rods in tne assembly with other assumptions as in the design basis analysis. The realistic analysis also assumes damage to the 56 peripheral rods with fuel rod gap activities taken from Table 11.1-12. The design basis and conservative cases use fifth percentile meteorology while the realistic case assumes fiftieth percentile meteorology. The mathematical models used in the analysis are described below: The atmospheric dispersion factors used in the analysis, a. which are based on meteorological conditions assumed present during the course of the accident, are calculated according to the model described in Subsection 2.3.4. 15.7-7 Revision 33 4/81
NUREC-0612 PART II RESPONSE MIDLAND 1&2-PSAR APMDmIX A N b. The thyroid inhalation dose, skin, and whole-body gamma immersion dose to an individual exposed at the exclusion area boundary and outer boundary of the low population (~gj zone (LPZ) are analyzed using the models described in Appendix 150. The potential radiological consequences of a postulated fuel handling accident occurring in the auxiliary building have been conservatively analyzed using the assumptions and models described. The skin and whole-body gamma dose due to immersion and the thyroid dose due to inhalation have been analyzed for the O to 2 hour release period at the exclusion area boundary and at the LPZ outer boundary. The resultant doses are listed in Table 15.7-6 for a postulated fuel handling accident in the auxiltary building and in Table 15.7-7 for a postulated fuel handling accident inside containment, and offsite doses are well below the guidelines of 10 CFR 100. The control room area ventilation system is designed to isolate and pressurize the control room upon the detection of high radiation from the radiation monitors located in the outside air intake of the control room. For this reason, the radiaticn doses to control room personnel are considered to be insignificant and are not evaluated. The detailed design of the control room ventilation system is given in'Section 6.4 and Subsection 9.4.1. 15.7.5 SPENT FUEL CASK DROP ACCIDENT IN j As discussed in Subsection 9.1.4.3.10, the Midland spent fuel l34 cask handling concept incorporates a single-failureproof overhead crano handling system, with which a cask drop accident is not 29 credible. t O V p. -- s 15.7-8 Revision 34 6/81 u r
I l mmzG-0612 PART II RESPONSI APPENDIX A MIDLAND 1&2-FSAR i TABLE 15.7-5 ("h y) PARAMETERS USED IN EVALUATING t THE RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT Design Basis Conservative Realistic Parameter Value Value Value Source Data Power level, MWt 2,552 2,552 2,552 Radial peaking factor 1.65 1.65 1.00 Earliest time after shutdown 100 100 100 that fuel handling begins, hr Number of failed rods 208 56 56' Fission products con-Table Table Table tained in the fuel 11.1-11 11.1-11 11.1-11 assembly, table number Fraction of fission Reg Guide Reg Guide Table (~1roduct gases contained 1.25 1.25 11.1-12 \\,ln the gap region of Assumptions Assumptions the fuel rods, % Activity Release Data i Fraction of gap 100 100 100 cctivity released, % Pool decontamination 1 1 1 factor for noble' gases Effective pool 100 100 100 dscontamination factor for iodine Iodine chemical form released from dcmaged assembly, % Elemental iodine 75 75 75 Organic iodine 25 25 25 Containment volume, ft 1.67E+6 1.67E+6 1.67E+6 l30 3 (sheet 1) Revision 30 10/80
m2EG-0612 PART II IIS?CNSI MIDLAND 1&2-FSAR AP'E3 DIX A (O) TABLE 15.7-5 (continued) l Design Basis Conservative Realistic Parameter Value Value Value Fraction of the con-0.1 0.1 0.1 tainment volume used for mixing for release through purge isolation valve Fraction of the con-8.38E-3 8.38E-3 8.38E-3 tainment activity released to the environment before the purge valve is closed Fraction of the con-0.5 0.5 0.5 30 tainment volume used to take credit-for mixing for release through equipment ' hatch /} Maximum ventilation 4,000 4,000 4,000 (,,/ flowrate through the equipment hatch, cfm Activity released from damaged assembly to the pool water, Ci Isotope I-131 4.83E+4 1.30E+4 1.00E+3 I-132 2.94E+4 7.92E+3 1.95E+1 1-133 3.78E+3 1.02E+3 8.48 I-134 0.0 0.0 0.0 I-135 3.43 9.23E-1 2.53E-3 Xe-131m 3.93E+2 1.06E+2 5.28E+1 'Xe-133m 1.04E+3 2.80E+2 2.18E+1 Xe-133 6.85E+4 1.84E+4 4.01E+3 l30 Xe-135m 0.31 8.40E-2 2.29E-4 Xe-135 1.55E+2 4.17E+1 2.01E-1 . I-mi (sheet 2) \\/ Revision 33 4/81
NUREG-0612 ? ART II RESPCNS: MIDLAND 1&2-FSAR APPENDLT A TABLE 15.7-5 (continued) a 9 Design Basis Conservative Realistic
- arameter Value Value Value_
Kr-85m 2.34E-3 6.29E-4 5.62E-6 Kr-85 2.01E+3 5.41E+2 9.15E+2 Kr-87 0.0 0.0 0.0 Kr-88 7.24E-7 1.95E-7 0.0 ...ergency fuel handling 0.99 0.99 0.99 l33 .rea exhaust charcoal filter iodine efficiency"8 Dispersion Data Distance to exclusion 500 500 500 area boundary, meters Distance to LPZ outer 1,600 1,600 1,600 boundary, meters ) Atmospheric dispersion () 3 factors, s/m L i Exclusion area boundary 0 to 2 hours 4.49E-4 4.49E-4 7.63E-5 l18 2 LPZ outer boundary 0 to 2 hours 8.49E-5 8.49E-5 1.09E-5 l18 "' Filter efficiencies are in accordance with NRC Regulatory l33 Guide 1.52. l30 OU (sheet 3; thviriori 2t j 4 ff:1 v'
NUREG-0612 MIDLAND 1&2-FSAR p 1,z TABLE 15.7-7 RADIOLOGICAL CONSEQUENCES OF A POSTULATED FUEL HANDLING ACCIDENT INSIDE CONTAINMENT Design Basis Conservative Realistic Rne. ult Value Value Value Exclusion Area Boundary Dore (0-2 hours), rem Thyroid 2.13E0 5.75E-1 7.22E-3 Skin 7.30E-1 1.96E-1 9.76E-3 30 Whole-body gamma 2.70E-1 7.26E-2 2.65E-3 LPZ Outer Boundary Dose (duration), rem Thyroid 4.04E-1 1.09E-1 1.03E-3 Skin 1.38E-1 3.71E-2 1.39E-3 30 iole-body gamma 5.10E-2 1.37E-2 3.79E-4 0 Revision 30 10/80 l
MIDLAND 1&2-FSAR NUREG-0612 PART II IESPONSE APPENDIX A cerr :pending t: pplicabic plant technical opccificati0n Or a ce per refueling cycle. O \\_/ 12.3.4. 6 High Range Containment Radiation Monitor A high rar.ge ontainment area radiation monitoring c
- ability, j
covering accid t radiation levels, is provided i e radiation monitoring system RMS) design as part of Midla accident j monitoring instrume tion (AMI). This cap ity provides the i control room operator 'th a direct indica on of the magnitude l of an accident in the co inment and potential radiological ~ s hazard it presents to the p lic. The high range containment radia 'on monitor is a Class lE device with control room indication th safety-related display console and continuous tren 'ng capab' ity on the AMI safety parameter display panel, edundant mon ors, qualified to the 30 requirements of Regula ry Guide 1 89 as cussed in Appendix 3A, are pro ded for each unit. Th e monitors cover a /hr with an energy response ' thin 20% from range of 1 to 10 7 l 80 kev through 2 V. For further information, ref to Tables 7.5-3 nd 12.3-2. .i Because e high range containment area radiation monito prim ly serves an accident monitoring function, in lieu o a pe onnel protection function, the design criteria specified neral plant area monitoring in Subsections 12.3.4.1.1 and j 12.3.1.1.2 arc not applicabic. I 12.3.4.2 Airborne Radioactivity Monitoring l 19 Airborne radioactivity monitoring is provided in compliance with 10 CFR 20 and Regulatory Guides 8.2 and 8.8. The' purpose of airborne radioactivity monitors is to monitor the air within an l19 enclosure by either direct measurement of the enclosure atmosphere or the exhaust air from the enclosure. Monitors 130-indicate and record the levels of airborne radioactivity and, if 19 abnormal levels occur, actuate alarms. Local alarms are provided to alert personnel to areas where airborne radioactivity is at or above the selected setpoint level to ensure that personnel are not subjected to airborne radioactivity above the limits of l 30 10 CFR 20. The monitors provide a continuous record of airborne radioactivity levels through the plant radiation monitoring 39 system minicomputer which will aid operating personnel in maintaining airborne radioactivity at the lowest practicable level. 12.3.4.2.1 Criteria for Selection of Airborne Radioactivity Monitors {j} The criteria for determining the type of airborne radioactivity s_ monitor are based upon the nature and type of radioactive
- 19 releases expected, and the location being monitored.
Revision 30 10/80 12.3-23
NUREG-0612 MIDLAND 1&2-PSAR TART II RESPONS A!?ENDIX A Where ingestion of radioactive airborne materials by plant p ernonnr1 in a ponr:i1 ility, combinatinn particulatn / halogen l ". lonitors are used to analyrc, record, and alarm should the (,,) ( 'iadioactivity approach the limits established by 10 CFR 20. Where detecting leakage is the primary concern, monitors are l provided to detect the highest activity isotope expected to be l present due to the leak. Where quick system isolation for l 19 protective purposes is the major objective, gamma detectors installed directly in the air stream being monitored are used. l30 In the case of containment airborne radioactivity monitors which are used to detect leakage from the reactor coolant pressure boundary, the guidance of Regulatory Guide 1.45 is followed, as discussed in Appendix 3A. 12.3.4.2.2 Criteria for Airborne Radioactivity Monitor Locations The criteria for locating airborne radioactivity monitors are dependent upon the point of leakage, the ability to identify the source of radioactivity so that corrective action may be performed, and whether personnel may be exposed to the airborne l19 radioactivity. Airborne radioactivity monitors sample operating areas a. for which there is a reasonable potential for airborne radioactivity. (A N_,/ b. Areas not normally accessible are monitored prior to personnel entry with portable monitors or samplers depending upon the potential for airborne radioactivity and work to be performed in the area. Portable airborne radiation monitors are provided to supplement installed airborne monitors. These monitors plug into the radiation monitoring system and function as installed units. Portable monitors are used in areas where work l being performed produces a high potential for airborne radioactivity, c. Areas containing processes which, in the event of major ,9 leakage, could result in concentrations within the plant approaching the limits established by 10 CFR 20 for plant workers will be monitored. d. Outside air intake ducts for the control room area will be monitored to measure possible introduction of radioactivity into the control room to ensure habitability of those areas requiring personnel l occupancy for safe chutdown. ,"s q, Revision 30* 12.3-24 10/80
NiiREG-0612 MIDLAND 1&2-FSAR TART II RESPONS ATPENDIX A 12.3.4.2.3 System Description (Airborne Radioactivity Monitoring) (Q 12.3.4.2.3.1 Introduction The airborne radioactivity monitors feed into the plant radioactivity monitoring system described in Subsection 11.5.2. 39 Two types of fixed instrumentation are used for monitoring l l airborne radioactivity, offline and inline. The offline system extracts a sample from the ventilation stream or area and l transports that sample to the radiation monitor which contains the specified equipment to detect particulates, halogens, and/or 19 noble gases. The inline airborne monitors are mounted within ventilation ducts and measure direct gamma radiation of the air l 30 l stream. l 12.3.4.2.3.2 Sampling Criteria l l The sampling system for the offline particulate / halogen / noble gas 19 monitors is designed and installed in accordance with the i American National Standard ANSI N13.1-1969 guide to sampling of l airborne radioactive materials. In cases where the sample 39 l te:aperature may affect detection devices, sample coolers, j temperature alanas, and automatic isolation may be provided. Par i monitors that may be subject to sample condensation, sample 30 conditioning and drain taps are provided to minimize the effects l 19 of condensation. 7_ 12.3.4.2.3.3 Detection Criteria l For offline monitoring of radioactive particulates and noble l gases, beta sensitive scintillation detectors are utilized to sense radioactivity. Use of beta sensitive devices serves to minimize l the effects of background radiation and, consequently, obtain a 30 l lower minimum detectable concentration. For halogen (iodine) l monitoring, charcoal filters and gamma scintillation detectors i are used. To be responsive to particulate, halogen, and gaseous activity in 99 l l ventilation ducts, the inline monitors utilize directionally i l sensitive g-13 tube detectors. These detectors are sensitive to 30 gamma radiation, i Where safety-related functions are required, redundant detectors 19 l are provided. 12.3.4.2.3.4 Instrumentation Criteria Instrumentation necessary to indicate, alarm, and perform control l19 functions will be provided to complete the monitoring system. (n) v 12.3-25 Revision 33 4/81
MIDLAND 1&2-FSAR NUREG-0612 PART II RESPONSE APPENDIX A Since radioactive concentrations may vary substantially, wide range instruments are utilized. bachmonitorisprovidedwithadownscalealarmtoindicate channel failure or loss of power. The observed activity levels from each monitor are indicated locally at the local control unit and in the control room via a CRT display. In the case of safety-related monitors, dedicated 19 indicators are provided in the main control room on the cafety-related display panels. Alarms are indicated both audibly l32 cnd visually on or near the local control unit and on the CRT I display in the control room. The activity levels are 1 19 continuously stored by the system minicomputer. Power for the non-class lE airborne radiation monitors is supplied from a diesel generator backed, non-Class lE, 120 Vac and 480 Vac power cource. Power to the non-Class lE airborne radioactivity monitors is tripped out upo'n either a LOCA trip signal or a loss 32 of offsite power. Power can be restored to the airborne radioactivity monitor.v after the trip by using administrative controls to connect the airborne radioactivity monitors to the Class lE bus network. Power for the Class lE monitors is supplied from a Class lE power source. 12.3.4.2.3.5 Sensitivities 19 The criteria for location of the monitors consider dilution Jactors in complying with the as low as practicable limitations s -of 10 CFR 20. For monitor sensitivities, see Table 12.3-3. 12.3.4.2.3.6 Calibration and Testing Airborne monitors are calibrated by the manufacturer for at least l19 the principal radionuclides listed in Table 12.3-3. The mEnufacturer's calibration standards are traceable to National Bureau of Standards and are accurate to at least 5%. The source datector geometry during this primary calibration is identical to the sample detector geometry in actual use. Secondary standards counted in reproducible geometry during the primary calibration are supplied with each airborne monitor. Secondary calibration standards may be different from the primary standard if it is 37 impractical to maintain the primary standard at the plant. The accondary calibration source set is portable, has a half-life greater than 10 years, and provides a radiation field of sufficient intensity to permit checking of all calibration 19 adjustments for each detector, including linearity. Each airborne monitor is calibrated periodically using the secondary l30 redienuclide standards. Monitors in use are calibrated at the 19 most conservative frequency corresponding to applicable plant technical specifications or at least once per refueling cycle. l30 ( The count rate response of each airborne monitor to a remotely 19 t,)ositionable check source (supplied with each monitor) is recorded by the manufacturer after the primary calibration, again Revision 37 12.3-26 9/81
NUREG-0612 MIDLAND 1&2-FSAR PART 11 RESPONS APPE!QIX A after installation, and periodically during system operation to 1 30 ] ensure proper functioning of the monitors. Recorded data showing check source response is maintained for 39 each monitor. Instrument responses falling outside of normal operating limits are investigated as to cause, and appropriate measures are taken to restore proper functioning. Following repairs or modifications, the monitors are recalibrated at the plant with the secondary radionuclide standards. Decay curves are provided for the sources to permit correction for source decay. jg 12.3.4.2.4 Airborne Radioactivity Monitor Locations 12.3.4.2.4.1 Introduction The specific type of monitoring system to be employed and the particular operating characteristics are included in the discussion of the particular area monitored. In general, an offline airborne monitoring station consists of monitors, a local control unit, and a positive displacement pump which draws a continuous cample of known volume and flowrate of the area to be monitored. The sample is drawn past detectors for jg the type of monitoring desired. Particulate monitors employ a 3 fixed filter (paper) and a beta scintillation detector. Halogen s._/ (iodine) monitors employ a cnarcoal filter and a gamma scintillation detector. Gaseous monitors employ a fixed volume and a beta scintillation detector. Solenoid operated valves which are operable from the control room are provided at the 30 sample chamber of each monitor -to permit purging of the sample chamber to facilitate background activity checks. In a combined monitor, the particulate monitor comes first, the iodine second, and the gas third. A particular monitoring station may employ 19 any combination of the three types of monitors for a single pump. Output from the detectors is processed in the local control unit and passed on to the radiation monitoring system minicomputer. An inline airborne monitor consists of a directionally sensitive 30 g-m tube detector and a local control unit. Safety-related inline airborne monitors also have a display and control station on the RMS safety-related display console in the control room. The local readout instrumentation consists of a digital display l 19 and associated circuitry as required to convert the pulse rate to a digital signal output suitable for indication, recording, and i 19 alarm trip circuits. High and high-high radioactivity level alarms are provided for each channel. l 30 I The sensitivities are given in Table 12.3-3. Initial alarm points are based on the most restrictive isotopes which are 19 () expected to be present and concentration levels as defined in Revision 37 12.3-27 9/81
EUIEG-0612 MIDLAND 1&2-FSAR PART II RESPONSE APPENDIX A Table I, Column I of Appendix B to 10 CFR 20 considering /~' dilution. Alarm sei.loints may be adjusted based on operating (_,hxperience. 12.3.4.2.4.2 Spent Fuel Pool Vent Monitoring System The spent fuel pool vent monitoring system is Class lE, Seismic 99 Category I, and completely redundant. The system utilizes inline detectors to continuously monitor the spent fuel pool exhaust cctivity upstream of the HVAC filters. A total of four independent monitors are provided, two in each train of the spent fuel pool exhaust system. The monitoring system provides an indication of activity released incident to a fuel handling accident. The spent fuel pool may also contain low level radioactivity due to partial mixing with the reactor coolant system during each refueling. This low level radioactivity will also be detected by the spent fuel' process monitor i 30 (Subsection 11.5.2.1.11). 19 The spent fuel pool vent monitoring system triggers high and high-high alarms. The high-alarm indicates when the radioactivity in the spent fuel pool vent is at area limits. I 32 f Receipt of a high-high alarm triggers the FPVAS which in turn terminates normal fuel building ventilation and shifts exhaust flow to the emergency fuel pool exhaust system. The lnstrumentation and control provided to initiate this action are 19 fescribed in Subsection 7.3.3.2.8. Details of the spent fuel pool vent monitors are given in Table 12.3-3. In addition to the spent fuel pool vent monitoring system, airborne activity monitors monitor the spent fuel pool area-30 atmosphere for particulate and iodine em.ivity (see Subsection 12.3.4.2.4.3). 119 12.3.4.2.4.3 Spent Fuel Pool Airborne Monitoring System Tve airborne monitoring stations are provided to monitor the 30 epent fuel pool area for airborne particulate and iodine activity. These are offline type monitoring stations and are provided for plant personnel protection. The spent fuel pool area airborne monitors will trigger high and high-high alarms. High alarms will alert plant personnel to 19 increasing levels of radioactivity. High-high alarms will indicate when either airborne particulate or iodine levels are at l MPC as defined in 10 CFR 20, Appendix B, Table I. Additional l details of the spent fuel pool area airborne monitors are given in Table 12.3-3. l Revision 37 12.3-28 9/81 i
MIDLAND 1&2-FSAR NUREG-0612 PART 11 RESPONS' ^ 12. 3. 4. 2. 4. <l Control Room Ventilation Monitoring System The cunt.Lol room makeup air monitoring system is a Class lE, seismic category I system and is completely redundant. It is denigned to continually monitor tne makeup supply air of the normal heating, ventilation, and air conditioning system for the 19 purpose of operator protection. The makeup air supply for the control room comes from outside air. A total of four independent monitors are provided, two in each train of the control room makeup air ducting. The monitoring system utilizes inline detectors located upstream of the HVAC filters. In the event 1 30 that radioactivity is detected in the intake air to the control the nyctem triggers a signal to the control room isolation
- room, system (CRIS) which isolates the makeup air supply.
The CRIS is described in Subsection 7.3.3.2.9. The system also triggers high alarms that alert the operators to increauing activity levels in the makeup air. Details of the radiation monitor are listed in 19 Table 12.3-3. In addition to the fixed monitors in the control room makeup air system, portable airborne activity monitors can be located in the control room to monitor for particulate and iodine activity. These portable monitors are non-Class lE because they have no safety system interface. The portable monitors are further 37 described in Subsection 12.3.4.2.4.6. 19 ("} 12.3.4.2.4.5 Containment Air Monitor k_/ The containment air monitoring system continuously monitors the containment atmosphere for airborne gaseous, particulate, and iodine radioactivity. The system consists of four noble gas channels, two particulate channels, and two iodine monitoring channels. Functionally, the system provides input to the reactor building isolation system (RBIS-II), detects leakage from the reactor coolant pressure boundary, and is used to alert plant 30 personnel entering or working in containment of increasing or abnormally high airborne radioactivity levels. The noble gas radioactivity monitoring channels of this system are Class lE, seismic Category I, and completely redundant. These monitors input to RBIS-II and trigger :ontainment isolation systems upon detection of high-high radioactivity levels. The noble gas monitor also provides indirect indication of reactor coclant pressure boundary leakage and alarms on increasing levels l37 of gaseous radioactivity in excess of the rate of change setpoint specified in Subsection 5.2.5.3.3. The particulate and iodine radioactivity monitoring channels of 30 this system are located on a separate monitoring skid. These monitoring channels are seismic Category I. Where these channels interface with a noble gas monitoring channel, the interface piping and seals are appropriately qualified to ensure that the e fs Revision 37 12.3-29 9/81 ..... ~. _ ... ~
mRIG-0612 MIDLA!JD lt=2-FSAR PART II RIS?ONSI l A?'E MIX A particulate and iodine ch uncls will not degrade the performance 39 f^30 ! the Clash lE uvble gau munitor. (s,)The particulate radioactivity r onitor 'provides indirect 30 indication of reactor coolant pressure boundary leakage and alarms on increaning levels of particulate radioactivity in i 37 excess of the rate of change setpoint specified in Subsection 5.2.5.3.3. Particulate radioactivity monitoring for leak detection is utilized in accordance with the requirements of Regulatory Guide 1.45. 30 Refer to Section 5.2.5 for additional information on reactor coolant pressure boundary leak detection systems. The containment atmosphere will be sampled by drawing air from the containment operating floor through the sample tubing to the 19 detectors located in the auxiliary building. The design of this system is shown in Figures 6.2-119, 6.2-120, and 11.5-1. The l 34 containment monitoring system, including particulate, iodine, and l 30 noble gas monitoring channels, provides plant operating personnel the capability of assessing the levels of airborne radLoactive contamination inside the containment prior to entry into the containment. The system will also provide information concerning the necessity of purging the containment and the length of purging required to allow personnel entry. The system is capable of providing: 19 /~') Information relative to variations in general airborne Q a. (s,/ radiation background v b. Data to augment personnel monito-ing measurements c. Warning signals, should the airborne radioactivity levels reach a level in excess of the limits specified in Appendix B, Table I, of 10 CFR 20. (The reactor building HVAC system is designed to recirculate the air 30 in the containment, thereby providing a uniform mixing cf the containment atmosphere.) Containment airborne radioactivity alarms will be generated for the following functions: High alarm to alert operators to increasing levels of a. 39 gaseous, particulate, or iodine activity inside containment b. High rate o'f change alarm on gaseous or particulate activity to alert operators to a possible leak inside containment 30 High-high alarm on gaseous activity to trigger c. containment isolation through the reactor building 19 r~s isolation system (RBIS-II, see Subsection 7.3.3.2.2) \\ ,s) r ,l Q,/ Revision 37
- 12.3-30 9/81
E,1EG-0612 MIDLAND 1&2-FSAR ?.t.RT II RES?CNSE AITENDIX A l Detailu of the containment monitors are given in Table 12.3-3. l 19 l /N U 12.3.4.2.4.6 Portable Monitors l Portable airborne activity monitors are provided to allow periodic localized monitoring of specific air volumes of interest independent of the fixed monitor systems. The monitors are used 39 to verify that airborne activity levels within the plant operating spaces are within allowable limits and also to verify the proper operation of fixed monitor systems. l Portable monitors are similar to the fixed offline airborne 1 37 activity monitors described in Subsection 12.3.4.2.4.1 and are supplied as particulate / iodine and particulate / iodine / gaseous l monitors. These monitors will plug into the plant radiation l monitoring system through receptacles located throughout the plant. l Portable airborne activity monitors will be used to monitor the l local areas where there is a possibility of airborne 19 radioactivity. Maintenance on radioactive systems, re fueling, anticipated operational occurrences, and accidents involving the spread of airborne radioactivity will be monitored locally using the portable monitors. Periodic grab samples for particulate and iodine will be taken throughout the plant and analyzed to ensure that the fixed monitors are operating properly. All monitors /' ) will be calibrated using radioactive sources traceable to (m/ National Bureau of Standards' standard sources. Details of the portable monitors are listed in Table 12.3-3. 12.3.4.2.4.7 Hefueling Canal Exhaust Monitors The refueling canal exhaust monitoring system is Class lE, seismic Category I, and completely redundant. The system, which is only functional during fuel handling operations, utilizes inline radiation detectors (g-m tubes) to monitor the refueling canal exhaust upstream of the HVAC filters. Two independent monitors are provided per unit. The monitors will be located on the exhaust system duct riser along the centerline of the fuel pool. From this location, che monitors will have "line of sight" l detection capability as vell as the ability to detect any 30 radioactive gas that enters the refueling canal exhaust system ductwork. The refueling canal exhaust monitoring system provides an l indication of activity released incident to a fuel handling accident in containment. The radiation monitoring system includes two radiation alarm setpoints. A high alarm setpoint d1erts personnel to increasing levels of radiation in the refueling canal area. Receipt of a high-high alarm triggers y RBIS-II which actuates containment isolation systems, thereby v) terminating normal refueling canal exhaust. During modes of Revision 37 12.3-31 9/81
mmEG-0612 MIDt.AND 142-PSAR TART II RESPONSE APEDmIX A plant operation other tnan fuel handling, electrical power to these monitors is turned off and the monitor's trip nignal to { (
- RBIS-II is bypassed.
Details of the radiation monitor are listed j I~) k / n Table 12.3-3. \\- i 12.3.4.2.5 Safety Evaluation The following monitors are located upstream of filters and, therefore, are effective for monitoring in-plant airborne radioactivity levels, a. Containment air monitors b. Spent fuel pcol vent monitors 19 c. Control room supply system monitors d. Spent fuel pool airborne monitors e. Containment refueling canal exhaust monitors i 30 The proccus radiation monitors for monitoring plant effluents are discussed in Section 11.5. Corresponding ratios among the various species of the airborne radioactivity will be periodically determined by laboratory analysis so that the leakage of halogen and gaseous radioactivity o can be accounted for. !9, In the event that radioactivity is detected in the auxiliary building via airborne or area monitors, portable airborne monitors will be used to monitor the local areas where there is a likelihood of excessive airborne radioactivity. Maintenance on the radioactive systems, refueling, anticipated operational occurrences, and accidents involving the spread of airborne radioactivity will be monitored locally using the portable monitors. on the basis of the above discussion, the airborne radioactivity monitoring system is adequate and sufficient to ensure personnel protection from airborne radioactivity. The system will provide indication to the operator that airborne radioactivity exists. The exact location of the airborne radioactivity can then be found by using portable airborne activity monitors that can be 39 manually connected at various selected sample points in the building. Use of these sample points will direct health physics personnel to the particular area of concern. The exact location of the airborne radioactivity can then be found by sampling the particular subcompartments in that area. The combination of the area and airborne radioactivity monitoring 19 systems, in conjunction with administrative controls restricting and limiting personnel access, standard health physica practices, (^') ventilation flow patterns throughout the plant, plant equipment g js QJ Revision 37 12.3-32 9/81
HUREG-0612 MIDLAND 1&2-FSAR ? ART II RESPONSE APMDOIX A l a'/ou t, minimization of sources in radiation Zones II and III, jg (~' and control of perudnnel access to areas, is sufficient to ensure \\, that airborne radtoactivity levels are safe in terms of the required duration of personnel access throughout all areas of the plant. A general review of these concepts follows: a. Equipment location is such that highly radioactive piping and equipment are located in radiation Zone IV and V areas to which entry is limited by administrative l 19 control. Radiation Zone II and III areas do not contain piping and components which would result in significant airborne radioactivity sources. This reduces the possibility of airborne radioactivity exposure to occupants of radiation Zone II and III areas where general entry is permitted. b. Air flow patterns are consistent with the basic ventilation design criteria of the plant. Clean filtered outside air is supplied to Zone II areas (corridors, clean areas); these areas are exhausted by drawing air into rooms and areas of successively higher potential for airborne contamination. Air flow is such that air flow reversal or exfiltration from potentially contaminated areas is precluded. This ventilation arrangement eliminates the possibility of personnel exposure to airborne radioactivity in continuous occupancy areas such as radiation Zone II areas. /'_,i x/ c. High radiation areas (radiation Zone V areas) where general area dose levels exceed 100 mrem /hr are kept restricted and conspicuously posted in accordance with 10 CFR 20. These areas are not normally entered. Personnel access to these areas is controlled by the administrative methods outlined in Subsection 12.5.3. 19 d. Health physics programs are discussed in Section 12.5. O Revision 37 12.3-33 9/81
waEG-0612 R3cponses to NRC Quastions ? ART II RESPONSE Midlcnd 1&2 ^?MDOU ^ ~} Question 312.44 (15.7.4) '~' Your response to request 312.28 in FSAR Revision 8 is insufficient to enable us to determine whether the consequences of a fuel handling accident would be acceptably mitigated. We require that this event be promptly detected, that the containment be automatically isolated, and that the offsite radiological consequences remain within 25% of the guideline values of 10 CFR Part 100. Provide an analysis which utilizes appropriate plant detection and isolation capability and which 16 conservatively allows for any mixing of the activity within containment, prior to release, in order to show that your design meets the above criteria.
Response
In response to Regulatory Guide 1.13, it was clarified that 56 peripheral fuel pins are assumed to be damaged during a fuel handling accident. This assumption was acceptable to the NRC staff as contained in their letter dated September 29, 1976. However, the containment purge air and refueling canal exhaust system has been modified such that the system exhaust,is not routed directly to the environment during refueling. Analysis of a fuel handling accident in containment considering these exhaust fI'N system modifications shows that the radiological consequences of 30 s,) the accident will remain well within the NRC required 25% of s 10 CFR 100 guideline values even for fuel handling accidents involving damage to all 208 fuel pins of a fuel assembly. The following sections have been revised in response to this i 39 questions i 30 a. Subsection 12.3.4.2 has been revised to include the 1 33 addition of redundant safety grade radiation monitors to the refueling canal exhaust system to ensure prompt 30 detection of a fuel handling accident in containment. b. Subsection 9.4.9 has been revised to reflect the 1 33 modifications made to the purge air and refueling canal exhaust system to ensure adequate time to isolate direct 30 containment discharges to the atmosphere to minimize l radiological consequences of a fuel handling accident. c. Subsection 15.7.4 has been revised to present the l 33 results of the analysis on the radiological consequences of a fuel handling accident in containment considering l the recent exhaust system modifications and assuming 30 i damage to all 208 fuel pins of a fuel assembly for the l design base. 1 & ' C Revision 39 Q&R 15.7-3 11/81
NUREG-0612 Racponses to NPC Questions
- * " ^
Midland 1&2 O d. Subsection 7.3.3.2.2 has been revised to reflect the 1 33 3 interface between the radiation monitors added to the purge air and refueling canal exhaust system and the 30 reactor building isolation system (RBIS-II). e. The responses to Questions 312.29 and 312.40 have been 139 revised to reference the response to this question. l30 i 1 i 1 I O D F I. 1 s O 3 Revision 39 Q&R 15.7-4 11/81
!RGIC-0612 FART II RESPONSE MIDLA::D 1&2-FSAR APPENDIX A rx 9.4.1.4 Tests and Inspectiona ) (O Ductworkistestedforleakageduring installation. A system air balance test and an adjustment to design conditions are conducted in the course of the preoperational test program. Each systen is operated and tested initially with regard to flowpaths, flow capacity, and mechanical operability. An abstract of the preoperational testing is provided in Chaptor 14 Tne system continues to be proved operable during normal plant operations. 9.4.9 RCACTOR DUILDI!!G AIR PURIFICATIO:1 MID CLCKiUP SYSTEM The reactor building air purification and cleanup system (RDAPCS) described in this subsection functions to provide a suitable atmosphere for operating personnel inside the containment during both normal and refueling operations by removing fission products from the containmenc atmosphere while maintaining radioactive discharges to the environment at a minimum. The reactor building heating, ventilating, and air conditioning system is described in Subsection 9.4.6. 9.4.9.1 Design Dases e' (s) Criteria for the selection of design bases are stated in Subsection 1.1.2.2. Environmental design is discussed in Section ~, 3.11. 9.4.9.1.1 Safety Design Bases The nDAPCS has no safety design bases. 9.4.9.1.2 Power Generation Design Bases POWER GC:!CRATION DCSIG:I BASIS 0:1C - The RSAPCS is designed to control containment airborno fission products during normal plant operation and shutdown in order to allow containnent access while limiting personnel exposures to less than the dose linits of 10 CFR 20 for occupational exposure and while limiting exposures resulting frca radioactive releases to the environnent to the cuideline values of 10 CFR 50, Appendix I. 9.4.9.1.3 Codes and Standards Codes and standards ~ applicable to the R3 APCS are listed in Table 3.2-1. The system is dosioned and conctructed in accordance with the requirements of N:CA, ARI, and S::AC::A standards. a Revision 2 9. 'l-3 G 12/77
NUREG-0612 PART II RESPONSE MIDLA!TD 152-FSAR APPENDIX A 9.u.9.2
System Description
9.4.9.2.1 General System Description The RDAPCS is shown schematically in Figure 9.4-10. The system consists of the refueling canal _ supply air system, purge air supply system, purge air and refueling canal exhaust system, air room supply and exhaust system, and the charcoal cleanup system. O l Revision 2 9.4-36a 12/77 4 .m
EmEG-0612 MIDLAND 1&2-FSAR P.iRT II RESPONSE APMDmIX A 9.4.9.2.2 Component Description ,A(,,) Design data for major components of the RBAPCS are listed in Table 9.4-13. REFUELING CANAL SUPPLY AIR SYSTEM - The refueling canal supply air system consists of two vaneaxial supply fans and associated instrumentation and controls. PURGE AIR SUPPLY SYSTEM - The purge air supply system consists of an outside air intake louver, an air operated inlet isolation damper, a roll filter, a face and bypass damper, a hot water heating coil, two 100% capacity vancaxial supply fans, containment ' isolation valves, and associated ductwork, instrumentation, and controls. The heating coil receives hot water from the plant heating system as discussed in Subsection 9.4.11. PURGE AIR AND REFUELING CANAL EXHAUST SYSTEM - The purge air and refueling canal exhaust system consists of a refueling canal e xh.ius t fan, control dampers, two airborne radiation monitors, containment isolation valves, a roughing filter, a high ef ficiency particulate air (UEFA) filter, two 100% capacity 30 vanoaxial fanu (each with an air operated mcdulating damper) and associated ductwork, ins trume n ta tion, and controls. AIR ROOM SUPPLY AND EXHAUST SYSTEM - The air room supply system consists of an outside air intake louver, an air operated inlet x ) isolation damper, a roll filter, a face and bypass damper, a hot y water heating coil, a chilled water cooling coil, a centrifugal fan, containment isolation valves, and associated ductwork, instrumentation, and controls. The heating coil receives hot water f rom the plant heating system as discussed in Subsection 9.4.11. The cooling coil receives chilled water from the auxiliary building chilled water system as discussed in Subsection 9.2.9. The exhaust system consists of containment isolation valves, a roughing filter, an upstream HEPA filter, a carbon adsorber, a downstream HEPA filter, a vaneaxial fan, an air operated outlet isolation damper, and associated ductwork, instrumentation, and controls. CHARCOAL CLEANUP SYSTEM - The charcoal cleanup system consists of a roughing filter, an upstream HEPA filter, a carbon adsorber, a downstream HEPA filter, a vaneaxial fan, and associated ductwork, 32 instrumentation, and controls. 9.4.9.2.3 System Operation NORMAL OPERATION - The charcoal cleanup system is operated as needed prior to personnel access of the containment. The system recirculates and filters containment air in' order to control airborne fission products. OV 9.4-37 Revision 32 1/81
NUREG-0612 MIDLAND 1&2-FSAR PART II RESPONSE APPENDIX A (")TTheair'roomsupplyandexhaustsystemoperatestoreducethe \\, airborne radioactivity in the air room to permit air room access (.3 during plant operation in order to limit personnel exposure to ,p less than the dose limits of 10 CFR 20 while limiting exposures resulting from radioactive releases to the environment to the guideline values of 10 CFR 50, Appendix I. A room thermostat controls the flow of chilled water to the cooling coil and operates the face and bypass damper as required to maintain its setpoint and thus maintain the room ambient air temperature. The air room exhaust is directed to the reactor building ventilation stack. The air room supply and exhaust system also has a provision to function as a low volume purge during startup, normal plant operation, hot standby, or hot shutdown. 15 SHUTDOWN OPERATION - The purge air supply system, refueling canal supply air system, and the purge air and refueling canal exhaust system operate to ensure the health and safety of operating personnel while inside the containment. The purge air supply System provides one air Change per hour of outside air to the containment. The refueling canal supply air system is used in conjunction with the purge air and refueling canal exhaust system to prevent the accumulation of any water vapor or radioactive gaseous activity released from the surface of the refueling canal during refueling. The refueling canal supply fans provide a 16,000 cfm air flow across the refueling canal. During periods when refueling operations are not taking place, 20,000 cfm of air is exhausted from the refueling canal via ductwork along the (cs s_) perimeter of the refueling canal. Control dampers are aligned to 30 ,) ) direct this exhaust, via the refueling canal exhaust fan to the \\m,/ purge air and refueling canal exhaust ductwork where it mixes with 8,000 cfm of containment air drawn from the purge air exhaust penetration at elevation 688'-6". The air is exhausted by the purge air and refueling canal exhaust fan to the reactor building ventilation stack after having been drawn through a prefilter and a HEPA filter. This mode of operation is cesigned to reduce containment temperature and humidity during times when fuel is not being handled. During refueling operations, the 20,000 cfm of air exhausted from the refueling canal is redirected by control dampers, via the refueling canal exhaust fan, to a point north of the equipment hatch. It is then discharged into the contaimnent at elevation 690'-0" and is directed away from both the purge air 30 exhaust penetration and the equipment hatch. The control damper on the purge air exhaust damper would be in the full open position to accommodate the full 28,000 cfm capacity of the purge air and refueling canal exhaust fan. Redundant Class 1E radiation monitors are provided in the exhaust ductwork and are located such that a radicactive release may be detected both by line of sight from the point of release at the refuelino canal surface and in the ductwork as it passes by the monitor after being drawn from the refueling canal surface. Upon g y detecting a radioactive releace, these moultors activate the s ) 'v) (o \\ (,f 9.4-38 Revision 30 10/80 l'
Ni1 REG-0612 ?A.C II.uS?ONSE MIDLAND 1&2-FSAR AP?ENDIX A ,/ x) reactor building isolation signal (RBIS-II). Subsection L/ 7.3.3.2.2 provides a description of RBIS-II and Table 7.3-3 identifies the equipment activated by RBIS-II. Included in the equipment activated by RBIS-II are the purge air exhaust penetration isolation dampers. These dampers close approximately 20 seconds following a monitored release of radioactivity. This time includes a worst case detection time of 5 seconds, 0.125 milliseconds to activate RBIS-II, and a maximum time of 15 seconds for closing the isolation dampers. If drawn into the exhaust ductwork, airborne radioactivity would require an estimated 40 seconds to reach this penetration. This is based on a minimum of 100 feet between the point of exhaust and the penetration and on a transport velocity of 150 fpm. The estimated time for the radioactive air to reach the penetration, 30 by a direct path from the refueling canal, is 17 seconds. This is based on a minimum straight line distance of approximately 43 feet between the refueling canal and the penetration and a transport velocity of 150 fpm. This time is very conservative in that the refueling canal supply fans would create convective currents which would prohibit the radioactive air from traveling directly to the purge air exhaust. Also, the steam generator shielding walls prevent a direct "line of sight" distance between the fuel pool and the purge air exhaust and inhibit the flow of air from passing directly to the purge air exhaust. (n) Based on this it is evident that in the first case the purge air N/ exhaust penetration would be isolated before any airborne radicactivity passes through it. In the second case, air could pass through the exhaust ductwork for a maximum of 3 seconds (1,400 ft3) before all the isolation dampers fully close.
- Also, t 33 this air would be mixed with at least 10% of the containment volume before it is released.
It is evident from this analysis that little or no airborne radioactivity would be released in this manner, and that it would be well below 25% of the 10 CFR 100 values. Radioactive releases through the equipment hatch have also been evaluated and are discussed in Subsection 15.7.4.2. The reactor bu.' ding vent stack radiation monitor includes a 32 radioactive gas channel, a particulate channel, and a radioiodine channel. This monitor initiates the closure of the purge air supply and exhaust dampers and results in the trip of the purge 37 air supply and exhaust fans. Detection of high radioactivity is annunciated in the control room. The containment air monitor continuously monitors the radioactivity of the containment 32 atmosphere, with indicating and recording devices located in the co~ trol room. The system's primary function is to actuate reactor building isolation signal, type II (RBIS-II) and also to determine when the containment may be safely entered and when it must be evacuated. These monitors are further discussed in () Subsections 11.5.2 and 12.3.4. l32 \\J Revision 37 9.4-39 9/81 E
i MIDLAND 162-FSAR NUREG-0612 ? ART II RESPCNSE APPEIGI.T A Through a temperature controller, a temperature sensor located r~'fownstream of the purge nupply air unit modulates a face and ( 'c)ypass damper to maintain its setpoint. All air purification and leanup system equipment is operable from outside the containment. Indication is provided on an HVAC panel located in the auxiliary building for position of dampers and operating status of fans. In addition, for the hydrogen vent exhaust, air room exhaust, and charcoal cleanup systems, flow through the 32 filter units and temperatures upstream and downstream of the charcoal filters are also indicated on the HVAC panel. Local differential pressure is provided across all filters. Alarms for low flow through filter units, high temperatures upstream and downstream of the charcoal filters, and high differential pressure across the HEPA filters upstream of the charcoal filters cre provided on an HVAC control panel. Upon sensing low l discharge air temperature (below 40F), the thermostat located in the discharge duct of the purge air supply and air room supply shuts off its respective fan and provides an alarm at the local penel. A discussion of the RBAPCS containment isolation valves is provided in Subsection 6.2.4. 9.4.9.3 Safety Evaluation Inasmuch as the reactor building air purification and cleanup (}jrovided./~' ystem has no safety design bases, no safety evaluation is g 9.4.9.4 Tests and Innocctions Ductwork is tested for leakage during installation. A system air balance test and an adjustment to design conditions are conducted in the course of the preoperational test program. Each system is operated and tested initially with regard to flovpaths, flew capacity, and mechanical operability. An abstract of the preoperational testing is provided in Chapter 14. The system continues to be proved operable during nornial plant operations. Water coils are pneumatically tested under water or hydrostatically tested to assure leaktightness. HEPA filters are mrnufactured and tested prior to installation in accordance with MIL-F-51068 and NIL-F-51079 as modified by NRC Health and Safety l32 Information Issue 306. After installation, HEPA filters are tested in accordance with ANSI 510-1975, Testing of Nuclear Cleaning Systems. Carbon adsorbers are tested after installation in accordance with ANSI 510-1975. f'N .A.) rm. \\*) Revision 37 9.4-40 9/01 e
.___m._.. ..m m __m. i t 1 l ) l CONTROL OF IfEAVY LOADS AT NUCLEAR POWER PLANTS (NUREG-0612) i PART II RESPONSE I APPENDIX B i REACTOR VESSEL HEAD DROP ANALYSIS i } i 4 ) i l e 1 1 t } i i I E \\
NUREC-0612 PART II RESPONSE APPENDIX B v CONTENTS i Page 1. INTRODUCTION. 1-1 2. ANALYSIS OBJECTIVES AND CRITERIA. 2-1 3. DETERMINATION OF LOADS............ 3-1 3.1. Weight Calculation.. 3-1 3.2. Calculation of Maximum Drop Height 3-3 4. REACIOR VESSEL GEOMEIRY AND MATERIALS 4-1 5. UNIFORM LOAD ANALYSIS 5-1 5.1. Method of Analysis...... 5-1 5.2. Spring-Mass Model 5-2 C'1 5.3. Axisymmetric Model...................... 5-3 5.4. Summary of Results. 5-5 6. POINT LOAD ANALYSIS 6-1 6.1. Load Description.. 6-1 6.2. Method of Analysis....... 6-1 6.3. Results of Analysis 6-3 6.4. Summary of Results.............. 6-4 7.
SUMMARY
AND CONCLUSIONS 7-1 8. RECOMMENDATIONS 8-1 ATTACHMENTS A. Material Properties and Allowable Stresses A-1 B. References B-1 List of Tables Table [ ) 5-1. Stiffness Properties of Spring-Mass Model. 5-6 \\s/ 7-1. Analysis of Heavy Load Drops - Summary of Results. 7-1 - 111 - Babcock & Wilcox
NURIC-0612 P.*aT II 22S?ONSI A??ENDIX 3 List of Figures Figure Page 3-1. RV Head and Attachments With Lifting Equipment 3-4 4-1. Longitudinal Section of Reactor Vessel and Support Skirt 4-2 5-1. Spring-Mass Model....................... 5-7 5-2. Ax1 symmetric Finite Element Model of RV Shell, Lower Head and Skirt 5-8 5-3. Reduced Finite Element Model With Substructured RV and Lower Head... 5-9 5-4. Reactor Vessel Shell Super Element 5-10 5-5. Lower Head Supper Element 5-11 5-6. Location of Stress Classification Lines. 5-12 5-7. Critical Buckling Mode Shape 5-13 5-8. Vertical Component of Velocity 5-14 5-9. Displacement Time History at Core Flood Nozzle, Uniform Load Case. 5-15 5-10. Displacement Time History at Core Flood Nozzle, Uniform Load Case. 5-16 6-1. Postulated Point Load Case 6-5 6-2. Three-Dimensional Finite Element Model 6-6 6-3. Longitudinal Section 6-7 6-4. Model After Substructuring 6-8 6-5. Displacement Time History Near Core Flood Nozzle Vertical Component Point Load Case 6-9 6-6. Velocity Time History Near Core Flood Nozzle i Vertical Component Point Load Case 6-10 6-7. Displacement Time History Near Core Flood Nozzle Horizontal Component Point Loao Case 6-11 6-8. Support Skirt Detail 6-12 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 i l 1 l ( O\\O' 1 - iv - Babcock a.Wilcox
NUREG-0612 PART II RESPONSE APEiDIX 3 O 1. INTRODUCTION The heavy loads analysis is an analysis to determine the effects of dropping heavy loads on the reactor vessel (RV) as prescribed by NUREG-0612.I The maxi-mum load lif ted above the reactor vessel is the RV head with service structure. Therefore, this analysis is limited to determining the effects of dropping the RV head onto the reactor vessel. The objective of the analysis is to deter-mine an allowable height to which the RV head may be lifted above the RV and dropped without violating the requirements of NUREG-0612. The analysis is divided into two sections (1) a uniform load' drop in which the head uniformly impacts the reactor vessel flange and (2) a point load drop in which the head falls at an angle and impacts the reactor vessel at.a point. Finite element models of the reactor vessel and its support skirt were con-structed to determine an allowable drop height. For the uniform load case an axisynsnetric model was used, while a three-dimensional, half-symmetry model was required for the point load. 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 analy-sis was a nonlinear, dynamic transient analysis using the ANSYS finite element computer program.2 The solution was evaluated in terms of the resulting l stresses in accordance with NUREG-0612.1 l O 1-1 Babcock a,Wilcox
NUREG-0612 PART II RESPONSE APPENDIX B 2. ANALYSIS OBJECTIVES AND CRITERIA 1 The objective of this analysis is to satisfy the requirements of NUREG-0612 as setforth in the following paragraphs. General guidelines for the control of heavy loads are given in Chapter 5.1 of NUREG-0612.1 The guideline applicable to this analysis is 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. Regarding the evaluation criteria specified in Section 5.1 of NUREG-0612,1 the rules of the ASME Boiler and Pressure Vessel Code, Section III, Appendix F have been selected as an appropriate set of accept'ance criteria. 8 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 " limiting the consequences of the specified event. They are in-tended to assure that violation of the pressure retaining boundary will not occur in components or supports which are in compliance with these procedures." The effects of the heavy loads considered are analyzed to the stress limits of Appendix F,8 thus satisfying the requirements of the evaluation criteria of NUREG-0612, Section 5.1.1 Appendix A of NUREG-0612 contains guidelines for conducting an analysis of a heavy load drop. Section A-1 contains general guidelines which should be 2-1 Babcock & Wilcox
NUREG-0612 PART II P.ESPCNSE APPENDIX 3 considered, as appropriate, in any heavy load analysis. These guidelines are listed below, followed by a discussion of how each guideline applicable to the structural evaluation of the reactor vessel, following a drop of the RV head, is satisfied in this analysis. The general guidelines are 1. That the load is dropped in an orientation that causes the most severe con-sequences. 2. That fuel impacted is 100 hours subcritical (or whatever the minimum that is allowed in f acility technical specifications prior to fuel handling). 3. That the load may be dropped at any location in the crane travel area where movement is not restricted by mechanical stops or electrical interlocks. 4. That 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 (-3m) water gauge. 5. Analyses that rely on 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 re-view. X/Q values should be derived from analysis of onsite meteorological measurements based on 5% worst meteorological conditions. 6. Analyses should be based on an elastic-plastic curve that represents a true stress-strain relationship. 7. The analysis should postulate the " maximum damage" that could result.1.e., the analysis 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 borated water makeup of adequate concentration that is required to be available by the technical specifications. I V 2-2 Babcock & Wilcox
NUREG-0612 PART II RESPONSi APPENDIX 3 O 10. Credit may not be taken for equipment to operate, that may mitigate the effects of the load drop if the equipment is not required to be operable by the technical specifications when the load could be dropped. The following is a point-by-point discussion of how each applicable guideline above is considered in this analysis: 1. The analysis considers two orientations of the RV head as it impacts the The uniform ' oad analysis considers the head dropping reactor vessel. l without tilting and impacting the RV flange uniformly. The point load analysis considers the head impacting the RV flange at an angle. 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 Attachment A to this report for material properties). 7. The RV head and its attachments are considered rigid and all energy is absorbed by the reactor vessel. 8. The RV head, with attachments, is the maxim m load lifted over the re-actor vessel. 9. Not applicable. 10. Not applicable. Section A-2 of Appendix A, NUREG-0612,1 contains additional guidelines to be considered when the postulated load drop is the RV. head. These guidelines are as follows:
- 1.. Impact loads should include the weight of the reactor vessel.(RV) head assembly (including all appurtenances), the crane load block, and other lif ting apparatus (i.e., the strongback for a BWR).
2. All potential accident cases during the refueling operation. Areas of consideration as a minimum should be ["N (} Fall of the RV head from its maximum height while still on the guide i a. studs followed by impact with the RV flange. N-2-3 Babcock & Wilcox i
WJREG-0612 ? ART II RES?CNSE A??ENDIX 3 b. Fall of the RV head from its maximum height considering possible ob-jects of impact such as the guide studs, the RV flange, the steam dryer (BWR), or structures beneath the path of travel. c. Impact with the fueling cavity wall due to load swing with the subse-quent drop of the RV head due to lifting device or wire rope failure. 3. All cases which are to be considered should be analyzed in the actual medium present during the postulated accident, e.g., for a PWR prior to reassembly of the reactor, the fueling cavity is drained af ter the head engages the guide studs to allow for visual inspection of the reactor core control drive rods insertion into the head. During this phase it should be considered thdt the head will only fall through air, without any drag forces produced by a water environment. 4. In those nuclear steam supply systems where portions of the reactor inter-nals extend above the RV flange, the internals should be analyzed for buck-ling and resultant adverse effects due to the impact loading of the RV head. It should be demonstrated that the energy absorption characteris-i tics (causing buckling failure) of these internals should be such that resultant damage to the core assembly does not cause a condition beyord \\ j the acceptance criteria for this analysis. 5. Reactor vessel supports should be evaluated for the effects of the trans-mitted impact loads of the RV head. In the case of PWRs where the RV is supported at its nozzles, the effects of bending, shear, and circumferen-tial stresses on the nozzles should be examined. For BWRs the effects of these impact loads on the RV support skirt should be examined. 6. The RV head assembly should be considered rigid and not experience deforma-tion during impact with other components or structures, t l l This analysis is in compliance with these guidelines as discussed below. 1. The weight considered in the analysis includes the RV head and all attach-I ments and lifting apparatus as described in section 3 of this report. 2. a. The drop heights considered in the uniform load analysis exceed the re-quirements of this guideline. b. Since this analysis is concerned with the reactor vessel, possible objects of impact other than the RV are neglected in order to maxi-mize the effects of the load drop on the RV. c. See response to b above. {) 3. All phases of this analysis assume a medium of air during the load drop, maximizing the impact velocity. 2-4 Babcock & Wilcox
NUREG-0612 PART II RESPONSE APPENDIX B s 4 The scope of this analysis is limited to an evaluation of the reactor ves-sel and its supports. 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 on the analysis guidelines prescribed by NURZG-0612. In addition to these guidelines, the following points are noted con-carning the analysis: 1. The increase in the ultimate tensile strength of steel when subjected to impact loading with a high strain rate is conservatively neglected in this analysis. 2. The stress i-the support skirt due to the daad weight of the internals and water is considered negligible compared to the impact stresses. The dead weight of the RV is, however, included in the analysis. 3. No structural damping is included. Damping is considered to have a neg-ligible effect on the peak response due to impact loading. 4. The stiffness of the attached piping is neglected. O A (--) 2-5 Babcock & Wilcox e
NUREG-0612 PART II RESPONSE APPENDIX B l 3. DETERMINATION OF IDADS, 3.1. Weight Calculation The load considered in this analysis is the RV head assembly plus all attached handling equipment used in lifting the head. The RV head assembly includes the following: 1. Reactor vessel closure head (including service structure support flange and control rod mechanism housings). 2. Closure head studs, nuts, and washers. 3. Service structure. 4. Control rod drives. 5. Head and service structure fixed pendants. ( ) 6. Stud parking spacers. 7. Chain hoists (four). The following handling equipment is used in lifting the RV head assembly: 1. Head and internals handling fixture. 2. Internals handling extension. 3. Movable pendants (t' o). w 4. Slings (three). Figure 3-1 shows the RV head assembly with attached handling equipment. The total weight of the above items is calculated on the following pages. The weight of the crane load block is not included in these calculations. This weight, not rigidly attached to the RV head and service structure, would re-sult in a secondary impact on the RV, and was not included in the total weight for the initial impact analysis. ( ~6 3-1 Babcock & Wilcox
NL%EG-0612 PART II RESPCNSE APPENDIX S Slings (reference B&W drawing 1010438-00) Total length = 73'-6" 2-5/8in. diameter-A={(2.625)2 - 5.4119 in.2 Volume = (73.5) (
- I 9) = 2.76 ft' 2a Weight - 2.76 (490 PCF) = 1354 lb - use 1400 lb Internals Extension (reference B&W drawing 203256E-01)
Mark 222 (two) - 2 " x 8'-8" x l'-2" 2 Volume = 2 5 (8.67)(1.167) = 3.79 ft 3 Mark 223 (two) - 34" x 14" x 3b" 3 Volume = 2 x 34(14)(3 ) 3 1.93 ft 3 Mark 327A (two) - use Mark 223 volume V Pins (four) - length = 15k in., diameter = 6k in. A={(6.25)2 = 30.7 in.2 3 Volume = 4(30.7)(15.25) 1.08 ft = 3 Mark 225 (two) - 11't 11" x 1h" x 8 Volume - 2 x 11 x 11 x 1.5 - 0.21 ft Mark 227 (one) 0.06 ft' Volume = 6"(12")(1.5") 12 3 = Total volume of internals extension: 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 ft 3-2 Babcock & Wilcox
NUREG-0612 PART II RESPONSE APPENDIX B 4 Weight = (9.0)(490 PCF) = 4410 lb - use 4500 lb Head Assembly (reference equipment specification 1005162-00) Weight = 309,060 lb Handling Fixture (reference B&W drawing 167191E-00) i Weight = 13,000 lb Pendants (Two) (reference B&W drawing 150170C-03) Weight = 1400 lb Total Weight Head assembly = 309,060 Handling fixture 13,000 = Two pendants 1,400 = Three slings 1,400 = Internals extension = 4,500 329,360 lb - use 330,000 lb 3.2. Calculation of Maximum Drop Height y/ The maximtsn height to which the RV he.ad assembly can be lif ted above the RV flange was determined to establish an upper limit on the drop height to be considered. The top of the RV flange is at elevation 634'-0" and the top of the polar crane rail is at elevation 740'-0" (reference B&W drawing 35375F-04). The main hook on the crane extends 2'-1" below the top of the crane rail." neglecting the clevis to maximize the lift height. The distance from the crane hook to the bottom of the,RV head is 54'-8-1/8" (see Figure 3-1). This gives a maximtsu drop height of (740'-0")-(634'-0")-(2'-1")-(54'-8-1/8") = 49'-2-7/8".
- A maximum drop height of 50'-0" was used.
C/ 3-3 i
- iUREG-0612 PART II RES?C:35E 1
APPENDIX 3 Figure 3-1. RV Head and Attachments With Lifting Equipment POLAR CRANE HOOK v INTERNALS D6 = HANDLING EXTENSION T 6 i= U .a I l HEAD & INTERNALS =, HANDLING FIXTURE S ASSIMBLY(TRIPOD) m \\ Il l L' TURNBUCKLE HANDLING FIXTURE j i PENDANT SLING ASSEMBLY U i m- +w i m HANDLING FIXTURE CLOSURE HEAD f SLING SERVICE 4 STRUCTURE REACTOR VESSEL AllGNMENT STUD CLOSURE HEAD ,'p i, I l l l r [ Babcock s.Wilcox 3-4
NUREG-0612 PART II RESPONSE APPENDIX 3 4. REACTOR VESSEL GEOMETRY AND.W ERIALS The reactor vessels for Midland Units 1 and 2 are both skirt-supported. The materials are Reactor vessel A-508 Class 2 Support skirt SA-516 Gra'de 70 Lower head SA-533 Grade B, Class 1 The properties for these materials and alluwable stresses are tabulated in Attachment A. Dimensions of these components were obtained from the drawings listed as references 5 through 49. The pertinent dimensions are summarized in Figure 4-1. l l O l 4-1 Babcock & Wilcox l
NL"dEG-0612 PART II RESPONSE AP?ENDIX 3 n v Figure 4-1. Longitudinal Section of Reactor Vessel and Support Skirt .625' - 8'*8' 2 3/8* - 8'-7' l '-2* l'-101/4" I/4" y o q i a b b l_ 7'-0* .b 6 \\ SEAL LittE g 9 C A
- l'-4 1/16'
'\\ %6'-10 11/16*-- 7'-0 3/16' w NN ' o.19" h 84 i b i: 7 E = 5 l l' o l d b R F i h 7'-l 11/16* 85.69" E 8.7/16" b 5 "~ s T 5 i e 1 e b o + 4 k 'j 6 t k \\ g t \\ f l h b ~ a 7/16-t m i T d L 7 5' E A 5
- 3. L6
,_ h. 1 o "f i o 'l 7'-5 3/4' 7s 4-2 Babcock & Wilcox
5 NUREG-0612 PART II RESPONSE APPENDlX 3 1
- O 1
a 5. UNIFORM IDAD ANALYSIS ? j 5. -l. Method of Analysis The effect of the RV head and attachments impacting the reactor vessel is modeled as an initial velocity problem. The mass of the RV head is lumped, { at the top of the RV ledge and is given an initial velocity to represent the impact. The validity of this approach and the implementation of the method i in the ANSYS computer program was verifie.1 through test cases run on ANSYS 2 and checked against hand calculations. The analysis was conducted in a three-step procedure as follows: 1. A nine degree-of-freedom (DOF) nonlinear spring-mass model was first con-structed. This model was used to determine an approximate allowable drop Ib height by analyzing the model for a series of different drop heights. i i 2. An axisynametric model was then constructed to obtain a more accurate re-sponse. This model was analyzed for the nav4== allowable drop height indicated by the results of the 9 DOF model analysis. ~ l i 3. The analysis was further refined by accounting for the conservation of j momentum at impact. The initial velocity approach used in steps 1 and 2 conservatively assumed that immediately following impact the RV and the j RV head moved downward with a velocity equal to the velocity at which the i head impacted the RV. By applying the principle of the conservation of momentum, the initial velocity of the RV and head can be reduced. This I reduced velocity is proportional to the impact velocity by the ratio of j the mass of the head to the combined mass of the head plus the mass in-pacted by the head. Although approximate formulas are available for in-l cluding this effect 50 the ANSYS gap element can be used to more adequate-l ly simulate the transfer of momentum upon impact. The gap element was included in the axisynenetric model, which was then re-analyzed for the critical drop height. 5-1 Babcock & Wilcox e i, .,...,,,,-m.-- ,m y,%., ,,,,,.,--,-y--
NUREG-0612 PART II RESPONSE APPENDIX 3 5.2. Spring-Mass Modal _ The 9 DOF model shown in Figure 5-1 was constructed to determine an approximate critical drop height and to provide preliminary information about the behavior of the system being analyzed. The model is composed of nonlinear axial-force elements with distributed mass representing the reactor vessel and a concen-trated mass element at the top node to represaat the RV head. The section properties of the axial elements are calculated in Table 5-1 and the nonlinear material properties are given in Attactanent A. Plastic material behavior is represented by bilinear kinematic hardening. The ANSYS program uses the maximum octahedral shear stress (or maximum distortion energy) theory of failure,2 which is in accordance with the requirements of Section III, Appendix F, Paragraph F-1321.1 of reference 3. A frequency analysis of the model was performed to determine an appropriate integration time step. The ANSYS program uses a form of the Houbolt direct integration method for integrating the equations of motion. This method has inherent numerical damping and the integration time step must be chosen so that the modes of interest are not excessively damped. The results of the frequency analysis showed a fundamental frequency of 64 Ez and included a maxi-mum frequency of 4010 Hz. The relationship between the integration time step and the numerical damping is given in reference 2. 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 0.00001 second was used to start the transient analysis. Later in the analy-sis a time step of 0.00003 second was used, which accurately integrates the first five modes. This model was then used in a nonlinear dynamic analysis. The loads applied were the weight of the reactor vessel applied statically and the combined weight of the RV head and attachments applied dynamically. An initial veloc-ity was applied at the RV ledge to model the impact due to a 50-ft drop of the RV head. This is the height determined in section 3 to be the maximum possible drop height. The results of this analysis indicate that the stresses from a l 50-ft drop would exceed the allowable stresses tabulated in Attachment A for the l support skirt. f_\\ b t 5-2 Babcock s Wilcox
NUREG-0612 PART II RESPCNSE A??ENDIX B g The model was then analyzed for drop heights of 8, 9, 10, 11, 12, 14, and 20 ft. Examination of the stresses in the support skirt (SA-516 Grade 70) indi-cated a preliminary allowable drop height of approximately 10 ft, with com-pressive stress approaching the allowable stress of 49,000 psi. The spring-mass medel also revealed that yielding is confined to the support skirt area for drop heights in the range considered. This allowed most of the reactor vessel to be analyzed linearly in subsequent models. The analysis of the 10-ft drop also indicated a maximum reaction of 55 x los Ib, which was used to evaluate the potential for skirt buckling. 5.3. Axisymmetric Model An axisymmetric model of the RV and support skirt, including the lower head, was constructed to perform a more detailed analysis of the head drop. The finite element mesh of this model is shown in Figure 5-2. Since the analyses done on the 9 DOF model indicated that the reactor vessel did not yield, all of the RV between the RV ledge and the transition piece was substructured to reduce the size of the model. It was anticipated that the lower head would not be stressed beyond yield and it was also substructured. The reduced model V is shown in Figure 5-3. The finite element meshes of the super elements in-dicating the degrees-of-freedom retained in the model are shown in Figures 5-4 and 5-5. A frequency analysis was also performed on this model. A fundamental frequency of 87.6 Hz was obtained. Integration time step calculations are as follows: An integration time step of 0.00001 second accurately integrates a frequency of 1/(30)(0.00001) = 3333 Hz, which includes the first 45 modes of vibration. A time step of 0.00006 second accurately integrates a frequency of 1/(30)(0.00006) = 556 Hz, which includes the first 13 modes. Time steps between 0.00001 and 0.00006 second were selected for use in the transient analysis. The axisymmetric model was analyzed for a drop height of 10 ft using the same analysis procedure previously described for the spring-mass model. The results of this analysis showed stresses in the skirt that exceeded the allowable stresses of Attachment A. A refined analysis was then conducted with a gap element as described in sec-tion 5.1, item 3, to take advantage of the reduction in the velocity of the d head upon impace with the reactor vessel. The re-analysis of the 10-ft drop 5-3 Babcock & Wilcox
NUREG-0612 PART II RESPCNSE APPENDIX 3 O height with the gap element became unstable after approximately 0.001 second into the transient (see Figure 5-8). This was corrected by restarting the analysis with the gap element removed from a point before the solution had gone unstable. Two consecutive sets of diaplacement solutions were taken from the gav element analysis and applied as specified displacements in the first two load steps of the restart analysis. The solutions at time t = 0.00030 second and t = 0.00031 second were chosen because the velocities of the gap element and the top of the RV model had converged at this point (Figure 5-8) and the support skirt had not yet yielded. The most critically stressed region being near the top of the support skirt, the stress classification lines of Figure 5-6 were used in an ANSYS post-pro-cessing routine to extract stress data in a form convenient for interpretation to ASME Code criteria. The highest stress intensities are Stress Allowable intensity, Figure 5-6 (Attactsnent A), Stress category psi stress lines psi Primary membrane 49,600 2 through 7 49,000 Primary membrane 71,400 2 73,500 plus bending The primary membrane stress intensity of 49,600 psi is equivalent to 1.01 S,, and the region over which the stress intensity exceeds 1.0 S, is approximately 6 inches in length. Accordingly, the maximum membrane stress intensity may be classified as a local membrane stress intensity (Paragraph NB-3213.10, Section III, reference 3) which easily meets the allowable value of 1.5 S, or 73,500 psi. Typical displacement time histories are shown in Figures 5-9 and 5-10. These are for the vertical and radial directions, respectively, of the core flood nozzle locations. Maximum displac' aments of reactor vessel nozzles are as fol-lows: l l J 5-4 Babc0Ck & WilCOX l
NUREG-0612 ? ART II RESPCNSE AP?!NDIX 3 Downward Outward vertical radial displacement, displacement, Nozzle in. in. Core flood 0.509 0.033 Reactor coolant
- 0.510 0.044
- Both hot leg and cold leg.
An eigenvalue buckling analysis was also performed with this model to evaluate the possibility of the support skirt buckling under Lapact. The results of the analysis show a critical load of 751 x 10' lb for skirt buckling. The corresponding buckling mode is shown in Figure 5-7. The analysis of the 9 DOF 8 model for a 10-ft drop height indicated a maximum reaction force of 55 x 10 lb which is sufficiently below the critical load to preclude buckling. 5.4. Summary of Results Under the assumption of the RV head making uniform contact with the RV flange, ( it has been demonstrated that the KV head can be dropped from a height of 10 ft above the reactor vessel and satisfy the level D stress limits of the ASME Code, with a margin of (73,500 psi)/(71,400 psi) or 3% on primary stress. The maximum compressive load in the support skirt has been shown to be about 7% of the critical b'ckling load. u The core flood nozzles would deflect a maximum of 0.509 inch vertically down-ward and 0.033 inch radially outward. Similar displacements for the hot and cold leg reactor coolant nozzles are 0.510 inch vertical and 0.044 inch radial. The effect of these nozzle displacements on the attached piping is summarized in section 7. ( k,- s L 5-5 Babcock & Wilcox
O O O Table 5-1. Stiffness Properties of Spring-Mass Model
- Length, E, 10' Element R1, ft-in.
Ro, ft-in. Area, in.' ft Material psi 1 7-0 8-4 9288 22 A-508 C1. 2 29.9 2 7 3/16 8 3/16 6800 38 A-508 C1. 2 29.9 3 7 3/16 8 3/16 6800 24 A-508 C1. 2 29.9 4 7 3/16 8 3/16 6800 45 A-508 C1. 2 29.9 5 7 11/16 7 -'10-1/8 4766 84 A-508 C1. 2 29.9 6 7 11/16 7 1/8 4766 83.625 A-508 C1. 2 29.9 7 7 11/16 7 - 8.09 3591 12.375 A-508 C1. 2 29.9 8 7 - 2.7 7 - 5.9 1766 34.25 SA-516 Cr 70 27.9 9 7 3/4 7 3/4 1115 28.75 SA-516 Cr 70 27.9 Y Mass 2 RV head: = 854.92 lb-s /in. 38 / 3 8 3 2 RV: mass density = (490) Ib/ft 1/12 ft /in.8 1/386 s /in. - 0.0007346 lb-s'/in." e k lE aaM 5? = 8 G"? .Nt m Si m
NUREG-0612 PART II RESPONSE APPENDIX B Figure 5-1. Spring-Mass Model h t 2 2 ( CORE FLOOD N0ZZLES 3 ( INLET /0UTLET ll N0ZZLES s (0 U 5 O h -(6) ji M y M 6 t b 7 U s 8 4 9 9 Ni ,f U H lf If 1f H if Il 5-7 Babcock 4.Wilcox
NUREG-0612 PART II RESPCNSE APPENDIX 3 Figure 5-2. Axisymmetric Finite Element Model of RV Shell, Lower llead and Skirt Y ~~ CORE FLOOD N0ZZLES X INLET /0UTLET N0ZZLES CD
- \\
DD q h J f f v Babcock & Wilcox 5-8
NIREG-0612 PART II RESPONSE AP?ENDIX 3 Figure 5-3. Reduced Finite Element Model With Substructured RV and Lower Head i I O 9 l FOR DETAll 0F SKIRT llESH l SEE FIGURE 5.2 i 5-9 Babcock 2.Wilcox
NL' REG-0612 PART II RESPONSE APPENDIX B Figure 5-4. Reactor Vessel Shell Super Element _) Y 4b Z ik 4F INDICATES NODES AT WHICH X AND Y DEGREES OF FREEDON ARE RETAINED AS (( NASTER DEGREES OF FREEDON FOR THE O SUPERELEMENT O O () O O OE0 i l O 5-10 Babcock & Wilcox
raEc-0612 PART II RESPCNSE APPENDIX B Figure 5-5. Lower Head Super Element Y O INDICATES NODES AT WHICH X AND Y DEGREES OF FREEDON ARE RETAINED AS NASTER DEGREES OF FREEDON FOR THE SUPERELEMENT. 6 INDICATES NODES AT WHICH THE'Y DEGREE OF FREEDON ONLY IS RETAINED AS A MASTER DEGREE OF FREEDON FOR THE SUPERELEMENT. l l l l O 5-11 Babcock & Wilcox
l NUREG-0612 PART II RESPCNSE APENDIX B Figure 5-6. Location of Stress Classification Lines < V" P 9.2 11 10 6 3 <,r,r i f i l I l' l 5-12 Babcock & Wilcox 1 l
NUREG-0612 PART II RESPONSE AP?ENDIX 3 i l Figure 5-7. Critical Buckling Mode Shape 9 O d E d I rlO i 5-13' Babcock & Wilcox 1 t
NUREG-0612 PART II RESPONSE APDENDIX 3 Figure 5-8. Vertica: Component of Velocity 0 ) i -40 80 O E -120 CONVERGE 0 VELOCITIES g O 3 -1 60 E -200 N00E AT TOP OF R V. f LEDGE (POINT OF lhPACT) -240 i GAP ELEMENT NODE -280 Y 320 .00000 .00021 .00043 .00064 .00086 .00107 .00123 .00150 Time (sec) l l i 5-14 Babcock & Wilcox l
NUREG-0612 ? ART II RESPCNSE A??INDIX 3 Figure 5-9. DisplacementTimeHistok,fatCoreFloodNozzle (Vertical Component), Uniform Load Case .040 s .060 .160 d 0 Ee 3 .260 i 1 2 l Ei O l .360 \\ .460 560 .0000 .0040 0080 .0120 .0160 .0200 ' Time 1 5-15 Babcock & Wilcox
NUREG-0612 PART II RESPCNSE l APPENDIX 3 Figure 5-10. Displace:nent Time History at Core Flood Nozzle (Radial Component), Uniform Load Case .0380 I I .0280 I .0180 - 0 3 .0080 - 5? } .0020 i L .0120 i k .0220 i i .0320 .0000 .0040 .0080 .0120 .0160 .0200 ' Time O 5-16 Babcock & Wilcox
NUREG-0612 PART II RESPONSE APPENDIX 3 6. POINT IDAD ANALYSIS 6.1. Load Description The other possible type of contact the falling head would have with the reactor vessel is the point load case. In this case it was postulated that the fail-i ure mechanism would cause the head to fall obliquely to the plane of the RV flange causing contact over a small area of the flange. It was further assumed that this contact area is sufficiently small to idealize this condition as a I point load. Referring to Figure 3-1, one type of failure mechanism which could cause point loading would be the. failure of one of the three slings attaching the lif ting lugs on the RV head to the head and' internals handling fixture, followed by failure of the other two. If this failure were to take place, the head would have an initial rotation about some horizontal axis which could lead to the RV head position of Figure 6-1, which shows a lower bound dimension of 104 inches for a lif t height corresponding to a postulated 60-inch or 5-f t drop t height. That is, the worst case vertical point load is with the center of gravity directly over the contact point. The analysis asstaned a 60-inch free 4 fall with the head in a rotated position. The head could actually be raised a minimum of 104 inches or 8'-8" prior to failure before the amount of energy transferred to the reactor vessel upon impact would exceed that for the 5-ft point load analyzed in this report. 6.2. Method of Analysis In order to accurately model this system a three-dimensional,180* (half 4 symmetry) model of the RV was necessary (see Figures 6-2 and 6-3). ANSYS was chosen as the finite element computer code because of its non-linear, dynamic problem solving capability.2 The shell and lower head of the RV were modeled using three-dimensional iso- [ parametric solid elements. These eight-node elements have three translational l degrees of freedom at each node. They may be redefined as triangular solids 6-1 Babcock & Wilcox
i NUREG-0612 FART II RESPONSE APPENDIX 3 and/or tetrahedrons as needed. The support skirt was modeled as a uniform. 2-inch thick cylinder using plastic triangular shell elements. 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. Interfacing these two element types was accomplished by using the Zienkiewicz in-plane rotation-al stiffness option (as formulated for the shell elements). RV upper lateral supports are present only in the Midland units. 'Ihese sup-ports were modeled in the present analysis and thus, this analysis is not di-rectly applicable to any other design. They require a third element type, the three-dimensional interface element. This element has two nodes with 6 DOFs at each node. It models the gap (at 70F) present during head removal / replace-ment as well as the axial compressive stiffness should contact be made. There are 12 upper lateral supports around the RV, six in the 180' model. Interpretation of the preliminary uniform load drop case revealed no undesir-able effects (plasticity, buckling, et'c.) in either the RV shell or the lower head. For this reason, along with the expectation that the system would be-have similarly for the point load case, it was decided to substructure the model to increase its efficiency. Most of,the RV shell was substructured into one large super element (see Figure 6-4). Similarly, most of the lower head was substructured into a second super element. These four element types comprise the 180' model. For this model to represent the reactor vessel, the translational degrees of freedom normal to the plane of symmetry of nodes were constrained to allow no displacement. This reflec-tive symmetry decreased the complexity of the three-dimensional model while permitting the system to simulate the response of a point load, provided that the load is applied in.the plane of symmetry. In addition, the lower nodes of the skirt elements were constrained in all 6 DOFs, representing the fixity between the support skirt and the concrete floor of the reactor cavity. The drop of the head was modeled as an initial velocity problem using time steps commensurate with those of the uniform load analysis. In this, the mass of the head impacts the RV ledge with the velocity it would have obtained after falling 5 ft. The entire mass of the closure head was assumed to impact the RV at a point Y on the plane of syrsnetry. This differs slightly from the uniform load analysis 6-2 Babcock a Wilcox
NUREG-0612 PART II RES?CNSE AFFENDIX 3 in that no gap element was used to simulate the drop. This was done because the use of the gap element had a very marginal effect on the earlier results.. 6.3. Results of Analysis The free f all from 5 f t was monitored at six positions through the model. Referring to Figure 6-3, these six locations are 1. At the point of contact. 2. Near the core flood nozzles. 3. At the inlet / outlet nozzles. 4. Mid-height of the RV shell. 5. One-quarter height of the RV shell. 6. At the c :p of the support skirt. Representative plots of time history responses are shown in Figures 6-5, 6-6, and 6-7. Figure 6-5 shows the vertical displacement near the core flood noz-zies and shows a maximum displacemnt of -0.336 inch. Figure 6-6 shows the corresponding velocity of this location. Figure 6-7 shows the horizontal dis-placement near the core flood nozzles; it reaches a maximum of 0.477 inch. For the point load analysis these maximum displacements are clso assumed to apply to the reactor coolant nozzles. Further analysis was done to obtain stresses in the first 12 elements around the skirt. This represents the segment from O' to 45' (where 0* is directly under the load). Due to the element numbering scheme used, the 12 elements are numbered 1-4,13-16, and 25-28 (see Figure 6-8). The vertical 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 found to be as follows: Stress Allowable intensity, Element No. (Attachment A), Stress category psi (Figure 6-8) psi Primary membrane 45,025 1 49,000 Primary membrane 46,180 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 calcu-laced vertical stress component observed and assume this acts uniformly l 6-3 Babcock &)Milcox
NUREG-0612 PART II RESPONSE A?PENDIX 3 N around the skirt. This " reaction" was then compared to the critical buckling load calculated in the uniform load analysis. The maximtse vertical stress (at the bottom of the skirt elements) occurred in element 13 and was 46,165 psi. If this value was uniform around the skirt (with an area of 1115.27 in.2), it would represent a maximum reaction of 51.5 x 10' lb. This value is well below the critical buckling load of 751 x 10' lb. 6.4. Summarv of Results It has been shown that the worst case point load drop for a RV head lift height of 8'-8" results in a maximum primary membrane stress intensity of approximately 92% of the allowable, a maximum primary membrane plus primary bending stress in-tensity of approximately 63% of the allowable, and a buckling load that is slightly less than 7% of its critical value. The vertical displacement near the core flood nozzles is 0.336 inch vertically downward and 0.477 inch radially outward. These same displacements may also be conservatively applied to the reactor coolant nozzles. Nozzle displacements are discussed further in the next section. T C' O l l 6-4 Babcock & Wilcox \\ ~ l
NURec-0612 PART II RESPCNSE A?PENDIX 3 Figure 6-1. Postulated Point Load Case CENTER OF GRAVITY DIRECTLY ABOVE CONTACT POINT CLOSURE HEAD WITH SERVICE STRUCTURE (IN ROTATED POSITION BEFORE VERTICAL OROP) n 104" LIFT HEIGHT a-(BEFORE ROTATION) 3 60" DROP HEIGHT F e l I O i REACTOR I I l l VESSEL 1 1 I / i ) k O 6-5 Babcock s.Wilcox
NUREG-0612 PART II RESPONSE APPENDIX 3 Figure 6-2. Three-Dimensional Finite Element Model /~ / -s s y / N s y 7 / / N \\ l / \\ / _, / \\ / f N \\ / \\ r r- \\ / s / / N \\ \\ _ / 7 N g' / / N \\ 7 y \\ _ / /- m / / l \\ N \\ / \\ 7 ( O 6-6 Babcock & Wilcox
NURIG-0612 PART II RES?CNSE A??ENDIX 3 Figure 6-3. Longitudinal Section LOAD 4U CLOSURE FLANGE r CORE FLOOD N0ZZLER = INLET /0UTLET N0ZZLES O <w 0 1 a-TOP OF / t' SUPPORT SKIRT =<> 7)7 7?P 7 e RESPONSE MONITORING LOCATIONS O 6-7 Babcock & Wilcox
NUREG-0612 PART II RESPCNSE i A?'ENDIX B Figure 6-4. Model After Substructuring /- ^ N N /- w\\ q / \\ / \\ / L RV SHELL SUPERELEMENT o I 7 SUPPORT = SKIRT LOWER HEAD SUPERELEMENT O l 6-8 Babcock & Wilcox
NUREG-0612 PART II RES?CNSE APPENDIX 3 O Figure 6-5. Displacement Time History Near Core Flood Nozzle Vertical Component Point Load Case DISP .040 .000 .040 .080 .120 O .1 60 .200 .240 .280 .320 .360 .0000 .0029 .0057 0086 .0114 .0143 .0171 .0200 Time 6-9 Babcock & Wilcox
NUREG-0612 PART II RESPONSE APPENDIX B Figure 6-6. Velocity Time History Near Core Flood Nozzle Vertical Component Point Load Case DISP. 100.0 80.0 60.0 40.0 20.0 O .0 -20.0 -40.0 - 60.0 - l ( L -80.0 -100.0 .0000 .0029 .0057 0086 .0114 .0143 .0171 .0200 O Time 6-10 Babcock & Wilcox
- iUREG-0612 P.%RT II RESPONSE A?PENDIX 3 Figure 6-7.
Displacement Time History Near Core Flood Nozzle Horizontal Component Point Load Case OlSP .500 .450 l .400 . 3 50 .300 O .250 .200 .150 .100 l .050 .000 .0000 .0057 .0114 .0171 .0229 .0286 .0343 O Time 6-11 Babcock & Wilcox
NUREG-0612 PART II RESNNSE APPENDIX 3 Figure 6-8. Support Skirt Detail Z l LOAD l Y 1 r 45' l s' x 22 / 14 15 21 16 j7 IB 19 g g g @ O / 1 2 3 ELEMENT NUMBER NODE NUMBER O 6-12 Babcock & Wilcox
NUREG-0612 PART II RESPCNSE-A?PENDIX B 'N 7. SLHMARY AND CONCLUSIONS An analysis of heavy load drops on the reactor vessel has been performed to the guidelines of NUREG-0612.1 Considered in this analysis was the RV head falling without rotation from a level position and making uniform contact with the RV flange. Also considered was an obliquely falling RV head th'at would impact the RV flange at a single point with the center of gravity of the RV head and attachments directly above the point of contact. Results from these two loading cases are summarized in Table 7-1. Table 7-1. Analysis of Heavy Load Drops -- Summary of Results -~(J Uniform Point Acceptance Analysis consideration load load criterion RV head lift height, ft-in. 10-0 8-8 NA General prinary membrane stress, psi 45,025 49,000 Local primary membrane stress, psi 49,600 73,500 Primary membrane plus bending stress, psi 71,400 46,180 73,500 Compressive load in support skirt, 10' lb 55 51.5 751 Core flood nozzle deflections, in. Downward vertical 0.509 0.336 (a) Outward radial 0.033 0.477 (a) Reactor coolant 'ozzle deflections, in. n Downward vertical 0.510 0.336 (a), Outward radial 0.044 0.477 (a) (*}All requirements of NUREG-0612 have been met.1 By limiting RV nozzle de-flections of the core flood and hot leg lines so as to maintain the integ-rity of the decay heat removal system, it is felt that the intent of NUREG-0612 can also be satisfied. The core flood nozzle deflections must be assessed by Bechtel for loads in the core flood piping. B&W has shown, I by conservatively scaling hot leg seismic pipe loads by reactor coolant l nozzle deflections, that the deflections should be acceptable, although a specific analysis should be performed to verify this conclusion. 7-1 Babcock & Wilcox
l I
- UREG-0612 l
?.A2T II RES?ONSE A??ENDIX 3 l l l I 8. RECOMMENDATIONS i Based on results from the uniform and point load drop analyses, it is recom-mended that the reactor vessel head be lif ted a maxista of 8'-8" above the reactor vessel flange. i t i 4 t 4 i t jl t i d a l 8-1 Babcock & Wilcox
NUR2G-0612 ?>RT II RES?CNSE A??ENDIX 3 ATTAC1DiENT A Material Properties and Allowable Stresses A-1 Babcock & Wilcox
NUREG-0612 ? ART II RISPCN52 APENDIX 3 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 on the following 'pages. These curves, provided by reference 51, are based on tests performed at B&W's Alliance Research Center in 1976 and 1979. As required by Section III, Appendix F. Paragraph F-1321.1b,3 the curves are adjuated to correspond to the tabulated values of yield stress in the ASME Code. ANSYS uses the energy of distortion (maximum octahedral i shear stress) method in elastic-plastic analysis, which is accepted by the i ASME Code, Section III Appendix F. Paragraph F-1321.lc. 3 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 ad-justed to the values of corresponding true stress-strain curves. 1.1. A-508 Class 2 (Reactor Vessel) 3 From 1977 ASME Code: oy = 50,000 psi 5 E = 29.9 x 10 pai From stress-strain curves: c = 0.108 a = 79,249 psi y
29 0, = 0.001672 c
Stress. True stress, Strain, c e o(1+c) Yield 0.001672 50,000 50,084 Ultimate D.108 79.249 87,808 1.2. SA-516 Grade 70 (support Skirt) 8 From 1977 ASME Code: y = 38,000 psi o 8 E = 27.9 x 10 psi r A-2 Babcock & Wilcox e
- URIG-C612
?.*?.T II RES?CNSE A??EITDIX 3 From stress-strain curves: c = 0.158 0 -20F c = 0.152 @ 125F o = 85,552 @ -20F o = 76,160 @ 125F Linearly interpolate to 70F: c = 0.1543 @ 70F o = 79,722 @ 70F e, = y = 2, o, = 0.0013e2 38 oa
- Stress, True stress, Strain, e o
a(1+c) Yield 0.001362 38,000 38,052 Ultimate 0.1543 79,722 92,023 1.3. SA-533 Grade B (Lower Head) 3 From 1977 ASME Code: y = 50,000 psi o 8 E = 29.9 x 10 psi Since the lower head material does not yield in the analysis, no nonlinear material properties are calculated. 2. Code Allowable Stresses The stress limits imposed by the ASME Code, Appendix F,'are as follows: General primary membrane stress intensity (P,): 1.0 S, Local primary membrane stress intensity (P ): 1.5 S, t Primary membrane + primary bending s.:ress intensity (P +P ): 1.5 S, b The S,value is calculated as the larger of 1. 0.7 S 2. S + 1 (S S) y 3 u y Babcock & WilCOX A-3 i
.~ NU'tEG-0612 PART II RES?CNSE APPENDIX B s The following values of S and S are taken from Section III Appendix 1: 3 y y + 1 (S -S }e 0.7 Su, S 8, Psi S, psi u y Material u y 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: P +Pb L Pm (1.0 S ), PL (1.5 S ), (1.5 Sg), m Material psi psi psi A-508 Class 2 60,000 90,000 90,000 i i SA-533 Grade B 60,000 90,000 90,000 l SA-516 Grade 70 49,000 73,500 73,500 i i i a l l l O Babcock & Wilcox g_4 l
O O O 1 l Figure A-1. Stress-Strain Curve for SA-508 Class 2 at 73F l 80,000 l g/ uts l l 60,800 d ys E v N 40,000 j dys = 53223 PSI kts=19249 PSI d. 20,000 i t e a i i 1 0 05 .10 .15 .20 .25 Strain (in/in) cn t* R w kkh P E i3 5 M i E &~? 0 M"8 U tm 2 iii"
O O O Figure A-2. Stress-Strain Curve for SA-516 Crade 70 at -20F l 80,000 duts 60,000 C E d, y T N 40,000 m w 8K t; = dys= 52301 PSI duts= 85552 PSI 20,000 l 0 8 8 8 O .05 .10 .15 .20 .25 .30 cn$ STRAllt 8n N> Maa M g = ? n n~a wNG m M M
O O O Figure A-3. Stress-Strain Curve for SA-516 Crade 70 at 125F 80,000 .I i duts 60,000 n ( 0 0 5 40,000 d = 49260 PSI u ud ys duts = 76160 PSI Tu 20,000 0 e i i e i 0 .05 .10 .15 .20 .25 .30 i Strain (in./in.) .'? cr g R- @si N SU g o a m,i: G 2 5 M
NUREG-0612 PART II RES?ONSE A??ENDIX B i ATTACIDiENT B References l i l 1 I Babcock & Wilcox B-1
NUREG-0612 PART II RESPONSE APPENDIX 3 1 NUREG-0612 " Control of Heavy Loads at Nuclear Power Plants," 1980. 2 G. J. DeSalvo and J. A. Swanson, ANSYS Engineering Analysis System User's Manual, Vol. I, II, Swanson Analysis Systems. Inc., Houston, Pennsylvania. 3 ASME Boiler and Pressure Vessel Code, 1977 Edition. Polar Crane Clearance Sketch, Drawing No.150A11D, Harnischfeger Corp. s Material List, Head and Vessel, B&W Drawing No. 142166E-08. Material List, Head and Vessel, B&W Drawing No. 142140E-09. 7 Closure Head Flange, B&W Drawing No. 142151E-03. e Closure Head Subassembly, B&W Drawing No. 142153E-10. Closure Head Center Disc, B&W Drawing No. 142152E-05. l' Reactor Closure Head Details, B&W Specification Drawing No. 136206E-03. 11 Miscellaneous Closure Head Details, B&W Drawing No. 142154E-08. 12 Closure Head Assembly. B&W Drawing No. 149919E-04. 13 Reactor Closure Head Details, B&W Specification Drawing No. 136211E-04 1" Head and Internals Handling Fixture Assembly, B&W Drawing No. 203254E-04 15 Turnbuckle Pendant Assembly and Details, B&W Drawing No. 150170C-03. l' Handling Fixture Sling Material List, B&W Drawing No. 101043B-01. 17 Turnbuckle Pendant Assembly Material List, B&W Drawing No. 101041B-00. l' Handling Fixture Material List, B&W Drawing No. 101042B-03. l' Head and Internals Handling Fixture Details, B&W Drawing No. 203255E-04. 2o Head and Internals Handling Fixture Assembly, B&W Drawing No. 167191E-00. 21 Head and Internals Handling Fixture Details. B&W Drawing No. 167192E-00. 22 Internals Handling Extension, B&W Drawing No. 203256E-01. 2s Material List, Head and Vessel, B&W Drawing No. 142140E-09. 2" Reactor Vessel, B&W Drawing No. 090532B-03. 2s Reactor Vessel Arrangement, B&W Specification Drawing No. 136203E-02. 2s A Reactor Vessel Details. B&W Specification Drawing No. 1362G*E-02, Sheet 1. ! V Babcock & Wilcox B-2
o NUREG-0512 FART II RESPONSE APPENDIX B 17 Reactor Vessel Details, B&W Specification Drawing No. 136205E-03, sheet 2. 2e Reactor Vessel Arrangement. Longitudinal Section, B&W Drawing No. 142138E-14. 2s Reactor Vessel Sections Arrangement. B&W Drawing No. 142139E-09. 8' Upper Shell Assembly, B&W Drawing No. 142141E-11. 31 Shell Assembly and Head Details. B&W Drawing No. 142142E-04. 32 Vessel Head and Support Assembly and Details B&W Drawing No. 142148E-07. 33 Vessel Assembly and Final Machining, B&W Drawing No. 142150E-04. s" Reactor Building Functional Requirement Elevation Section X-X, B&W Draw-ing No. 35375F-04. 35 Material List Head and Vessel, B&W Drawing No. 142166E-08. Reactor Vessel, B&W Drawing No. 090554B-02. 37 Reactor Vess91 Arrangement, B&W Specification Drawing No. 136208E-02. 3' Reactor Vessel Details. B&W Specification Drawing No. 136209E-02, Sheet 1. 3' Reactor Vessel Details, B&W Specification Drawing No. 136210E 'J3, Sheet 2. "8 Reactor Vessel Arrangement, Longitudinal Section, B&W Drawing No. 142164E-12. "1 Vessel Head and Support Assembly and Details, B&W Drawing No. 142174E-00. 42 Vessel Assembly and Final Machining, B&W Drawing No. 142150E-04. "3 Reactor Vessel Sections Arrangement, B&W Drawing No. 142165E-10. Upper'Shell Assembly, B&W Drawing No. 142167E-07.
- s Shell Assembly and Head Details, B&W Drawing No. 142168E-05.
Vessel Assembly and Final Machining, B&W Drawing No. 142176E-04. "7 Miscellaneous Upper Shell Details, B&W Drawing No. 142172E-08.
- a Reactor Vessel Sections Arrangement, B&W Dra. wing No. 142165E-10.
Upper Shell Forging, B&W Drawing No. 104176D-00. so R. J. Roark and W. C. Young, Formulas for Stress and Strain, 5th Ed., McGraw-Hill (1975). [] J. D. Aadland to H. T. Harrison, Memorandum, " Stress-Strain Curves for 31 l V ? SA-516 Gr 70, SA-508 C1. 2, and SA-533 Gr. B," Babcock & Wilcox, l July 20, 1981. l l Babcock & Wilcox 3_3
.. _ _..__ _. ~ _.. _ _. _.. _.. _. _. _ f i^ 1-il ' j jf i. ? i l l l i j t' i s t i CONTROL OF HEAVY LOADS AT i NUCLEAR POWER PLANTS i (NUREG-0612) ,h PART II RESPONSE j. APPENDIX C 4 + SUPPLEMENTARY INFORMATION FOR LOAD DROP ANALYSIS, i o C, AUXILIARY BUILDING - EL 659'-0" li g i. I i i i 4 I X ) s F. d., I .s. -e l [ T-i g. i _ / t,-. P c J'r 'd \\ - [ -b i a
- lAy, 7'
'z (.- j r ft ' (r.iif f l ,k (. [n, R z, I' ', j,_ .kj. 2..
NUREG-0612 PART II RESPONSE APPENDIX C V ANALYSIS OF PLANT STRUCTURES AUXILIARY BUILDING REFUELING LEVEL (El 659'-0") Generic Letter 81-07, Attachment 4 Information 1. Initial Condition / Assumptions for Heavy Loads Reactor Coolant Heavy Load Pump Motor Small Snubber Large Snubber
- a. Weight, tons 50 1
6
- b. Impact Area, sq in 452.4 8.75 8.75
- c. Drop Height, in 6
6 6
- d. Drop Location Fig C-7 Fig C-7 Fig C-7 between Column between Column between Column Lines B/C &
Lines B/C & Lines B/C & 6.2/6.9 (also over the over the see Fig C-2) W36x245 beam W36x245 beam extending North extending North to South be-to South be- ~ (~h tween 6.6 & 6.9 tween 6.6 &,6.9 \\# (see Fig C-2) (see Fig C-2) e. Impact Limiter None applied None applied None applied
- f. Floor thickness, in 21 21 21
- g. Drag Forces None applied None applied None applied
- h. Load Combinations ( )
- i. Material Properties Structural Steel ASTM A 36 ASTM A 36 ASTM A 36 Fy = 36 ksi Fy = 36 ksi Fy = 36 ksi Reinforced Compressive Compressive Compressive Concrete Strength Strength Strength 28 day 4,000 psi 4,000 psi 4,000 psi Method of Analysis (
2. 3. Conclusion (See NUREG-0612, Section 5.1) Criteria III NA NA NA Criteria IV May limit safe Does not limit Does not limit shutdown function safe shutdown safe shutdown llks# function function
2 NOTES: ( Load combinations: load drop and dead load of structure with no concurrent live load (2) Analysis per Bechtel Power Corporation Topical Report, BC-TOP-9A, Revision 2, dated September 1974, Design of Structures for Missile Impact; accepted by the AEC as referenced in applications for construction permits and operating licenses - November 25, 1974; computer codes and experimental test data not used. s 6_,}}