ML20040D883
| ML20040D883 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 01/31/1982 |
| From: | STONE & WEBSTER ENGINEERING CORP. |
| To: | |
| Shared Package | |
| ML20040D878 | List: |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR ES-FC-81-04, ES-FC-81-4, NUDOCS 8202020347 | |
| Download: ML20040D883 (125) | |
Text
{{#Wiki_filter:_ _____ - ____ ______ ________ _ ___________ __ ______________ ____ _ _ ___ _ ____________ _______ I TABLE OF CONTENTS l TITLE PAGE NUMBER 1.0 PURPOSE 1 2.0 CRITERIA 1 f 2.1 General I 2.2 Referenced Documents 2.3 Method of Analysis 1 1 1 2.4 Structural Criteria 3 I 2.5 Mechanical & Electrical Criteria 2.6 Nuclear Technology Criteria 5 6 3.0 I RESPONSE TO ENCLOSURE 3 0F NRC LETTER DATED DECEMBER 22, 1980 (the following section numbers are from the above 8 referenced document.) 2.2 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING 9 SYSTEMS OPERATING IN THE VICINITY OF FUEL STORAGE POOLS 2.3 SPECIFIC REQUIREMENTS OF OVERHEAD HANDLING 14 SYSTEMS OPERATING IN THE CONTAINMENT 2.4 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING 20 SYSTEMS OPERATING IN PLANT AREAS CONTAINING I EQUIPMENT REQUIRED FOR REACTOR SHUTDOWN, CORE DECAY HEAT REMOVAL, OR SPENT FUEL POOL COOLING HAZARD ELIMINATION TABLES 4.0 PROPOSED CORRECTIVE ACTIONS 33 5.0 SCHEDULE Appendix A Decision Chart Appendix B Reference Drawings and Sketches Appendix C Information Requested in Attachment 1,
" Single Failure Proof Handling Systems" Appendix D Criticality Analysis Data Appendix E Radiological Analysis Data I
genera 2% I e
I I 1.0.0 PURPOSE NUREG 0612 dated July,1980, " Control of Heavy Loads at Nuclear Power I 1.1.0 Plants", requires utilities to review their overhead load handling equip-ment, systems, and procedures to preclude the possibility of a load drop accident, the consequences of which could affect the safety of the plant. I The purpose of this report is to evaluate the cranes and monorails at the Fort Calhoun Station per the criteria of NUREG 0612, its attachments, and the USNRC letter dated December 22, 1980. The evaluation is intended to I determine system compliance or non-compliance with the criteria of NUREG 0612. Based upon this evaluation, recommendations to upgrade the safety of heavy load handling at the Fort Calhoun Station are made. 2.0.0 CRITERIA 2.1.0 GENERAL 2.1.1 The following sections describe the specific criteria and approach utilized by the structural, mechanical, electrical, and nuclear techno-logy disciplines in evaluating the load handling systems at the Fort Calhoun Station. 2.2.0 REFERENCED DOCUMENTS , 2.2.1 The following referenced documents were used as the basis for this report: A) USNRC letter signed by Darrell G. Eisenhut dated December 22, 1980 w/ enclosures:
- 1) NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants" I 2) 3)
Staff Position " Interim Actions for Control of Heavy Loads"
" Request for Information on Control of Heavy Loads" B) USNRC letter signed by Darrell G. Eisenhut dated February 3, 1981 C) OPPD Responses to Section 2.1, dated June 22, 1981 and November 30, 1981 " Request for Additional Information en Control of Heavy Loads" D) Fort Calhoun FSAR, as ammended.
I E) Stone & Webster Topical Report, SWECO 7703 (Dated Sept, 1977)
"Missil Barrier Interaction" F) Nuclear Regulatory Guide 1.29 I G) Nuclear Regulatory Guide 1.4 i
j 2.3.0 METHOD OF ANALYSIS 2.3.1 A heavy load is defined by NL' REG 0612 as any load that weighs more than the combined weight of a single spent fuel bundle and its handling tool. l
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For the purpose of this analysis loads weighing more than 1500 lbs. are considered heavy loads. I I
I 2.3.2 The consequences of a heavy load drop were evaluated for the worst case of all identifiable loads carried by the following cranes: I A) Containment Polar Crane 1) 2) 130T Main Hook 10T Auxiliary Hook B) Auxiliary Building Crane I 1) 75T Main Hook
- 2) 10T Auxiliary Hook
. C) Deborating-Demineralizer Crane (Auxiliary Building) D) Concrete Slab Removal Crane #1 (Auxiliary Building) E) Concrete Slab Removal Crane #2 (Auxiliary Building) F) Waste Evaporator Equipment Handling Crane (Auxiliary Building) G) Intake Structure Overhead Crane 2.3.3 The evaluation is intended to determine compliance or non-compliance with the following criteria (NUREG 0612 Section 5.1): I. Any release of radioactive material that may result from damage to spent fuel based on calculations involving accidental dropping of a postulated heavy load, will produce doses that are I less than or equal to one-fourth the limits of 10 CFR Part 100 (i.e. 75 rem thyroid, 6.25 rem whole body). II. Damage to fuel and fuel storage racks, based on calculations involving accidental dropping of a postulated heavy load, will not result in a configuration of the fuel such that K ' " is larger than 0.95. III. Damage to the reactor vessel or the spent fuel pool, based on I calculations of damage following accidental dropping of a postulated heavy load, is limited so as not to result in water leakage that could uncover the fuel. (Makeup water provided to overcome leakage should be from a borated source of adequate concentration if the water lost is borated.) IV. Damage to equipment in redundant or dual safe shutdown paths, I based on calculations assuming the accidental dropping of a postulated heavy load, will be limited so as not to result in loss of required safe shutdown functions. 2.3.4 A " drop envelope" was developed for each crane based on the limits of hook travel and lifv_d loads. The drop envelope was then projected down through the structure; any safety related components (piping, wiring, equipment, reactor vessel, spent fuel pool, as defined by NUREG 0612) were located and identified. A decision chart (Appendix A) was then used as a guide in determining if the effects of a heavy load drop comply with the criteria of NUREG 0612. I I I I
I 1 2.4.0. STRUCTURAL CRITERIA l 2.4.1 An impact analysis, within the drop envelope, was performed to I determine the survival of a safety related system or component as defined by NUREG 0612. l 2.4.2 The following assumptions were made regarding the dropping of a heavy l load within the drop envelopes: A) The load is dropped in an orientation that causes the worst consequences. B) The load may be dropped at any location in the crane travel area, except where movement is restricted by mechanical stops, I C) electrical interlocks, or where physical interference is present. Loads were not analyzed if their load paths or drop consequences were included in other load analyses. Consideration was given to drag forces resulting from the I D) E) environment of the postulated accident. Lift height was assumed to be the maximum physically possible for the given load / crane combination. I F) G) Piping, wiring, equipment, etc. in the path of the dropped load was assumed to provide zero resistance to the dropped load. Any piping, wiring, equipment, etc. impacted by a dropped load I H) was assumed destroyed. Any piping, wiring, equipment, etc. shielded by a structure was assumed destroyed if the structure failed the impact analysis. I 2.4.3 Structural targets included in the analysis were determined by their relative proximity to critical systems and components as defined by mechanical, electrical, and nuclear technology criteria. 2.4.4 In drop envelopes with several structural targets, the analysis was conducted until a target failure was indicated or until all targets were shown to survive the impact. 2.4.5 The following assumptions were made regarding structural analysis: I A) The analysis postulates the maximum potential damage of the structure (i.e., all energy is absorbed by the structure). B) I The analysis was based on an elastic plastic curve that represents a true stress strain relationship. C) Stresses and deflections in structural components prior to dropped load impact were assumed negligible. D) American Concrete Institute (ACI) load and under-capacity I factors were not considered in the impact analysis of concrete structures. I I I
I E) Compression steel was not considered in the determination of the plastic moment (Mp) in concrete slabs and beams. I F) Concrete slab and beam boundary conditions and reinforcing were based on available information and conservative assumptions. G) Concrete slabs and beams were assumed to deform plastically at I failure, and separate into segments at the yield lines. H) The reinforcing steel in concrete slabs and beams fully yields I along the yield lines at failure and developes the full plastic moment. I) At failure, elastic deformations are negligible compared to the I plastic deformations. 2.4.6 The method of analysis was as follows: 1 A) The basic kinetic energy at impact for the dropped loads was calculated by: KE = w(h 7
-h)2 where: KE= Kinetic energy at impact I w= weight of heavy load .'
h y = elevation of center of gravity of heavy load at release I h = elevati n f center f gravity f heavy load at impact 2 B) When applicable, the basic kinetic energy was reduced by:
- 1) Fluid drag forces on the dropped load when falling through a water environment.
- 2) Energy lost in previous impacts with structural targets.
I C) The energy required to cause failure of a structural component was determined by developing a force-deflection curve for the least force failure mechanism. This was accomplished by equating the external work done by a calculated failure force to the internal work done by the plastic moment. In equation form:
- 1) EW = P x d i I 2) IW = I (Mp x 0) 3)
4) IW = EW (P x d) = I (Mp x 0) I i I I I
I I where EW = External Work P = failure load I d = the unit deflection through which the failure load acts IW = Internal Work Mp = plastic moment of structure at yield lines I- 0 = the angles established by yield line geometry through which the plastic moment rotates. D) Equation (C-4) was solved for P for plausible failure mechanisms. I The miminum value P was then used to develco a force-deflection curve for the structural component under analysis. E) 'The deflection of a structural component at yield was determined using a cracked section analysis. _The component was assumed to fail when flexural deflection reached ten times its yield deflection. F) The area under the force deflection curve represents the total I ' energy absorbtion capacity of the structural component. This was compared with the maximum energy dropped load to determine survival or failure of the structure. I G) If the above bending analysis indicated survival of the structural component, a punching failure check was made based on Stone & Webster topical report SWEN 7703 (dated September, 1977), "Missle-Barrier Interaction,". 2.5.0 MECHANICAL & ELECTRICAL CRITERIA I 2.5.1 The safety related station mechanical and electrical systems / components were identified utilizing the criteria set forth in Reg. Guide 1.29 (Rev 3) and NUREG 0612. , Pursuant to NUREG 0612, post accident systems were not evaluated ! unless such systems serve a dual function in normal station l g operation. Specifically NUREG 0612 identifies the following ! g " exempt" systems (a) emergency core cooling, (b) post-accident containment heat removal and/or (c) post accident containment f - atmosphere cleanup. Such systems are exempt frou analysis since a load drop accident is , not assumed to occur in combination with other postulated accidents !g (i.e., pipe rupture at power etc.). l l l l I I I
I 2.5.2 The evaluation considered critical systems that are functional prior to the time the plant has attained a safe shutdown condition. It also considered systems which are required to maintain decay heat removal I during refueling operations. Systems or portions of systems which were evaluated are: I A) Reactor Coolant B) Chemical and Volume Control
- C) Shutdown Cooling Section of LPSI D) Spent Fuel Pool Cooling I
- E) Segments of the Raw Water, Component Cooling Water, and Auxiliary Feedwater required for:
- 1) residual heat removal from the reactor 1 2) cooling the spent fuel storage pool F) Portions of the Steam and Feed ater Systems (in excess of 2-1/2 I inches in diameter) extending irom and including the secondary side of steam generators up to and including the outermost contain-ment isolation valves, up to and including the first valve (includ-I ing a safety or relief valve) that is either normally closed or capable of automatic closure during all modes of normal reactor operation.
G) Systems or portions of systems that are required to supply fuel to emergency diesel generators. I H) Portions of Radioactive Waste Management systems whose postulated failure would result in conservatively calculated potential offsite doses (using Reg. Guide 1.4) that are more than the NUREG 0612 offsite limits. 2.5.3 Compliance with NUREG 0612 will be based on demonstrating: A) Survival of required systems (including the associated sub-systems) after load drop to attain safe shutdown condition from normal operation. B) Survival of required systems after load drop during refueling operation to provide for decay heat removal from the core. 2.6.0 NUCLEAR TECHNOLOGY CRITERIA 2.6.1 All analyses were completed in conformance to NUREG 0612 guidelines, procedures, and criteria.
- Systems considered for Decay Heat Removal during refueling operation.
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I I 2.6.2 Two potential consequences of heavy load drop accidents over spent fuel and new fuel were analyzed: A) Airborne radiation releases B) Criticality (K,gg <0.95) I 2.6.3 Analyses were conducted for two phases of shutdown: I A) Removal of the missile shields, RV head, and upper guide structure. For the shutdown mode, it was assumed that the spent fuel was subcritically cooled for 72 hours in borated water (FSAR minimum time from SCRAM to refueling operations). This assumption yields the worst case situation for an airborne I radioactive release. Analysis was also performed for fuel suberitically cooled for 240 hours. Past refuelings indicate this to be the average time from SCRAM to refueling operations. B) Replacement of upper guide structure, RV head, and missile shields. For the shutdown mode, it was assumed that new fuel I was loaded into the core. This assumption yields the worst case scenario for criticality. 2.6.4 General Assumptions A) Control assemblies and fuel assemblies cannot be driven through the bottom of the core due to the core support plate (2-inch I thick 304 stainless steel reinforced with support beams). Therefore, " crushing" and/or buckling of Control Element Assemblies (CEA's) and fuel assemblies could occur upon impact of the dropped RV head and UGS. B) Regarding any drop over the spent fuel, physical damage would be insufficient to produce water leakage which would uncover the I hot spent fuel. C) No CEA's can be ejected upward out of the core from any downward load impact. I I I I 'I I
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!I I \I i
- l. g Section 3.0 RESPONSE TO ENCLOSURE 3 I Sections 2.2, 2.3 and 2.4 l
l OF NRC LETTER DATED DECEMBER 22, 1980 1 !I lI 'I I I I I ~ I I I - --
I I 2.2 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN THE VICINITY OF FUEL STORAGE POOLS t NUREG 0612, Section 5.1.2, provides guidelines concerning the design and operation of load-handling systems in the vicinity of stored, spent . fuel. Information provided in response to this section should i I demonstrate that adequate measures have been taken to ensure that in this area, either the likelihood of a load drop which might damage spent fuel is extremely small, or that the estimated consequences of
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I such a drop will not exceed the limits set by the evaluation criteria of NUREG 0612, Section 5.1, Criteria I through III.
- 1. Identify by name, type, capacity and equipment designator, any I cranes physically capable (i.e., ignoring interlocks, moveable mechanical stops, or operating procedures) of carrying loads which could, if drapped, land or fall into the spent fuel pool.
RESPONSE: The only crane that carries a heavy load in the vicinity of the spent fuel pool is the auxiliary building crane. This crane is a double girder overhead bridge crane with a 75-ton I capacity main book and a 10-ton capacity auxilia y hook. The crane consists of Harnishfeger bridge girders and end trucks (100-ton capacity) retrofitted with an Ederer trolley and d y hoists. 2. Justify the exclusion of any cranes in this area from the above category by verifying that they are incapable of carrying heavy I loads or are permanently prevented from movement of the hook centerline closer than 15 feet to the pool boundary, or by providing a suitable analysis demonstrating that for any failure mode, no heavy load can fall into the fuel-storage pool. RESPONSE: No exclusions
- 3. Identify any cranes listed in 2.2-1, above, which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely small for all loads to be I carried and the basis for this evaluation (i.e., complete compliance with NUREG 0612, Section 5.1.6 or partial compliance supplemented by suitable alternative or additional design features). For each crane so evaluated, provide the I load-handling-system (i.e., crane-load-combination) information specified in Attachment 1.
I RESPONSE: The 75-ton capacity main hook of the Auxiliary Building Crane has been modified to meet the criteria of NUREG 0554. (see Appendix C for the information requested by Attachment 1 of I Enclosure 3 to USNRC letter of December 22, 1981). Since the 10-ton auxiliary hook of the subject crane is not single failure proof, it will be addressed separately to demonstrate compliance with NUREG 0612. I I
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- 4. For cranes identified in 2.2-1, above, not categorized according to 2.2-3, demonstrate that the criteria of NUREG 0612, Section I
5.1, are satisfied. Compliance with Criterion IV will be demonstrated in response to Section 2.4 of this request. With respect to Criteria I through III, provide a discussion of your evaluation of crane operation in the spent fuel area and your determination of compliance. This response should include the I following information for each crane: I
- a. Which alternatives (e.g., 2, 3, or 4) from those identified in NUREG 0612, Section 5.1.2 have been selected.
RESPONSE: Because the Auxiliary Building Crane has a single failure I proof main hook, and a non-single failure proof auxiliary hook, the response is based on a combination of Sections 5.1.6 and 5.1.2-4 from NUREG 0612.
- b. If alternative 2 or 3 is selected, discuss the crane motion limitation imposed by electrical interlocks or mechanical stops, and indicate the circumstances, if any, under which I these protective devices may be bypassed or removed. Discuss any administrative proedures invoked to ensure proper authorization of bypass er removal, and provide any related or proposed technical specification (operational and I surveillance) provided to ensure the operability of such electrical interlocks or mechanical stops.
RESPONSE: Not applicable
- c. Where reliance is placed on crane operational limitations I with respect to the time of the storage of certain quantities of spent fuel at specific post-irradiation decay times, provide present and/or proposed technical specifications and discuss administrative or physical controls provided to ensure that these assumptions remain valid.
RESPONSE: Not applicable
- d. Where reliance is placed on the physical location of specific fuel modules at certain post-irradiation decay times, provide present and/or proposed technical specifications and discuss I administrative or physical controls provided to ensure that these assumptions remain valid.
RESPONSE: Not applicable
- e. Analyses performed to demonstrate compliance with Criteria I I through III should conform to the guidelines of NUREG 0612, Appendix A. Justify any exception taken to these guidelines, and provide the specific information requested in Attachment 2, 3, or 4, as appropriate, for each analysis performed.
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I I RESPONSE: The travel of both the main hook and the auxiliary hook is prevented over the spent fuel pool by electrical interlocks. The area of hook travel with activated interlocks is shown in I Appendix B, Figure B-2 and Architectural drawings. With the interlocks active, both the auxiliary hook and the main book are restricted from traveling within l'-0" of the outer edge of I the spent fuel pool wall (6'-6" from inside edge). This is less than the 15 foot separation limit required by NUREG-0612, Section 5.1.2-2a. However, NUREG 0612 Section 5.2 page 5-10 states that the 15 foot separation limit is " based on the maximum dimensions of a cask to assure that in a cask tip, the center of gravity will not go beyond the edge of the spent fuel pool". Though the current separation limit (6'-6") does not meet the I NUREG-0612 criteria of 15 foot separation limit, subsequent discussion will justify the established 6'-6" separation limit. The interlocks can be by passed for the handling of certain I loads with approved written procedures from the Plant Review Committe (PRC). I Justification for Established Separation Limits: The load lifts handled by the crane can be divided into the following two categories:
- 1. Load lif ts/ movements at or outside the Crane Limits I The walls of the spent fuel pool are elevated 13'-6" above the surrounding floor level (finished floor elevation 1025'-0"). The limits of hook travel have I
been set at l'-0" outside the outer edge of the spent fuel pool wall. A dropped load that impacts the top of the spent fuel pool wall will have its center of gravity 1 foot beyond the outer edge of the wall (the center of I gravity would be directly below the hook centerline). This would result in an overturning moment which forces a tip of the load away from the spent fuel pool. I Dropped loads impacting the floor at elevation 1025"-0" would be prevented from tipping into the spent fuel pool by the elevated walls. I A load drop near the interlock limits nearest the spent fuel pool would impact the outer edge or face of the 4-foot thick concrete spent fuel pool wall. This may I cause a local spalling of concrete but the steel liner inside the pool will not be damaged. The intact liner will prevent any leakage from spent fuel pool that could uncover the stored spent fuel. I I
I I 2. Loads lifts / movements inside the Crane Limits: Fort Calhoun Technical Specification 2.11 and Operating I Instructions OI-HE-5 state that the Auxiliary Building Crane shall not be used to move any loads in the spent fuel pool without a written procedure. This procedure I should describe all the precautions and conditions and be approved by the Plant Review Committee. A key to bypass the electrical interlocks can be obtained from I the Shift Supervisor af ter the procedures are complete. Operating Instruction 01-HE-5 specifies that a certified /qualifiied Crane Supervisor is required to be in administrative control when the electrical interlocks are inoperative. At present, the following loads have been identified as I the heavy loads which are or will be moved over part of the spent fuel pool. The movement of these loads will require a written approved procedure to bypass the Auxiliary Building Crane electrical interlocks.
- a. Spent Fuel Shipping Cask and Spent Fuel Sample Cask I OPPD does not presently utilize or own a spent fuel sample cask and spent fuel shipping cask. Due to weight and safety considerations, the single failure proof main hook of the Auxiliary Building Crane would be used to move these casks. This I would preclude the possibility of a cask drop.
- b. Fuel Inspection Stand The fuel inspection stand is currently in the cask I
area of the spent fuel pool. This is a temporary installation. At present there are no plans to return the inspection stand to the spent fuel pool after removal. Since this is an infrequent load I lift and special precautions will be taken, the possibility of a load drop is extremely small.
- c. New Minimum Pitch Poison Design Spent Fuel Storage Racks.
The new spent fuel racks are expected to be installed in 1983 and this will be a one-time I lift. When installed, a specific procedure for removal of the existing racks and installation of the new racks will be written. This procedure will I outline all necessary precautions to make the possibility of a load drop extremely small. I I I
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- d. Spent Fuel Pool Gate I The non-single-failure-proof 10-ton auxiliary hook of the Auxiliary Building is used to move the spent fuel pool (SFP) gate.
I Maintenance Procedures MP-FH-16-1-1 and MP-FH-16-2-1 cover the removal and replacement of the spent fuel pool gate. These procedures require that the gate I be moved directly from its notch in the spent fuel pool wall to storage in the transfer canal (or vice-versa) without transport over irradiated fuel. This prevents the dropping of the SFP gate directly onto the spent fuel pool liner or racks. Therefore, if I dropped, the gate would fall either into the transfer canal, or into its notch in the SFP wall between the transfer canal and the spent fuel pool. Dropping the SFP gate into the transfer canal would not I damige spent fuel, since no spent fuel is stored in the canal. The bottom of the transfer canal consists of a stainless steel liner plate over a 12-foot thick coacrete mat on caissons. The worst postulated damage would be I the puncture of the stainless steel liner plate. Since the concrete behind the plate would retain its integrity, it is unlikely a leak rate would develop which would ex-ceed the borated water make-up capability and would have I s2gnificant adverse radiological impact. Nevertheless, should coolant be lost from the pool, the minimum coolant level attainable is the bottom of the SFP gate notch, I elevation 1008'-6". This level is above the top of the spent fuel stored in the racks. Therefore, there is no uncovering of the spent fuel. If the gate impacts the notch in the spent fuel pool wall causing damage to the wall, the resultant effects would be negligible. The SFP wall separates the spent I fuel pool from the transfer canal, both of which would be filled with coolant at the time the gate was moved. Therefore, there would be no loss of coolant from the spent fuel pool. The following scenario is assumed to describe the only SFP gate drop accident that could impact the I spent fuel racks:
- 1) The gate would have to be positioned over the I notch in the wall between the spent fuel pool and the transfer canal. The gate's orientation would have to be perpendicular to its normal I position (an unlikely situation).
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I I 2) A rigging failure occurs that causes a rotation of the gate as well as a gate drop.
- 3) The gate impacts the bottom of the notch, pivots at its base, and tips onto the spent fuel racks.
A structural analysis of the above accident scenario was performed. The analysis indicates local crushing I at the top of the spent fuel storage cells. However, the amount of this crushing is significantly less than the distance from the top of the spent fuel storage cell to the top of the spent fuel bundles (14 inches). I Therefore, damage to the spent fuel bundles would be avoided. I Since the spent fuel remains undamaged, Criteria I
& II of NUREG 0612 are satisfied. The crushing of the spent fuel racks absorbs the energy of the SFP gate drop thereby protecting the SFP liner. Since I there is no loss of coolant, Criteria III of NUREG 0612 is satisfied.
I 2.3 SPECIFIC REQUIR"MENTS OF OVERHEAD HANDLING SYSTEMS OPERATING IN THE CONTAINMENT NUREG 0612, Section 5.1.3, provides guidelines concerning the design and operation of load-handling systems in the vicinity of the reactor core. Information provided in response to this section should be sufficent to demonstrate that adequate measures have been taken to I ensure that in this area, either the likelihood of a load-drop which might damage spent fuel is extremely small, or that the estimated consequences of such a drop will not exceed the limits set by the evaluation criteria of NUREG 0612, Section 5.1, Criteria I through III.
- 1. Identify by name, type, capacity, and equipment designator, any crancs physically capable (i.e., taking no credit for any I interlocks or operating procedures) of carrying heavy loads over the reactor vessel.
I RESPONSE: The Containment Polar Crane is the only crane capable of carrying a heavy load over the reactor vessel. This is a double girder overhead bridge crane manufactured by Harnish-feger with a 130-ton main hook and a 10-ton auxiliary hook.
- 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 I 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.
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I I RESPONSE: No exclusions I 3. Identify any cranes listed in 2.3-1, above, which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely small for all loads to be carried and basis for this evaluation (i.e., complete compliance I with NUREG 0612, Section 5.1.6, or partial compliance supplemented by suitable alternative or additional design features). For each crane so evaluated, provide the load-handling-system (i.e., crane-load-combination) information specified in Attachment 1. RESPONSE: Not Applicable
- 4. For cranes identified in 2.3-1, above, not categorized according 1 to 2.3-3, demonstrate that the evaluation criteria of NUREG 0612, Section 5.1, are satisfied. Compliance with Criterion IV will be I 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 I following information for each crane:
- a. Where reliance is placed on the installation and use of I electrical interlocks or mechanical stops, indicate the circumstances under which these protective devices can be removed or bypassed and the administrative procedures invoked to ensure proper authorization of such action. Discuss any I related or proposed technical specification concerning the bypassing of such interlocks.
RESPONSE: Not Applicable
- b. Where reliance is placed on other, site-specific I considerations (e.g. , refueling sequencing), provide present or proposed technical specifications and discuss administrative or physical controls provided to ensure the continued validity of such considerations.
m RESPONSE: Not Applicable )g5 ! c. Analyses performed to demonstrate compliance with Criteria I through III should conform with the guidelines of NUREG 0612, l Appendix A. Justify any exception taken to these guidelines, l and provide the specific information requested in Attachment i 2, 3, or 4, as appropriate, for each analysis performed. I I . 'I I
1 I I RESPONSE: Criteria I The load lifted by the Polar Crane which has the greatest I potential for damage to the spent reactor core during refueling operations is the Reactor Vessel Closure Head with the lifting rig assembly attached. All other heavy load drop I accidents over the reactor core are encompassed by this radiological analysis. These include the Upper Guide Structure (UGS) and lifting rig and the concrete missile I shields and lifting rig. Appendix A.2.1). (This satisfies NUREG-0612, The bounding worst-case accident involves a radiological I release from a full core of spent fuel (133 fuel assemblies). The accident assumes a 36-foot drop of the RV Head and lift rig prior to refueling on to the upper guide I structure with all energy of the drop going to crush the core releasing all of the gaseous fission products available. The RV Head and lift assembly is assumed to impact at an angle onto the UGS such that maximum deformation damage is transmitted to the core fuel assemblies. Immediately after the crushing of the core, the gaseous I fission product activity would be released to the refueling water, and subsequently to the isolated recirculated and filtered containment atmosphere (this satisfies NUREG-0612 A.1.1 and A.2.2 considerations). I Specific information and assumptions used in the radiological analyses are documented in Appendix E, Table E.1. For an I isolated containment, the site doses are observed to be within one-fourth 10CFR Part 100 limits as shown in Table E.2, Appendix E (results are from DRAGON Code calculations). Table E.3, Appendix E documents that these calculations are I conservative for the postulated radiological release consequences. I To determine the effect from a less leak tight containment, Table E.4, Appendix E was derived and the results are plotted in Figure E.1. This information provides a basis for I estimating the accident consequences with relaxed containment leak rates and various fuel cooling times. The values presented in Table E.4 indicate that I iodine-filtration alone will not bring the potential radiological release from whole-core rupture within the one-fourth 10CFR Part 100 limits, even for the fuel which has I been cooled for 240 hours. Accordingly, containment isolation must be effective to limit the whole body (i.e. noble gas) release to achieve the reduced limits (of 10 CFR 100) specified by N1; REG 0612 (Appendix E). I I
1 m Under the assumption that containment isolation could not be achieved, the number of fuel assemblies which could be damaged - I and still meet the requirements of NUREG 0612 would be 26 and 65 for 72-hour and 240-hour cooled fuel respectively. These values were based on the assumption that the containment leak rate was 10 times that of the closed rate. With the equipment hatch door open, the number of fuel assemblies which could be I damaged and still meet the requirements of NUREG-0612 would be 1 and 2 for 72 hours and 240 hours cooled fuel respectively. I Various radiological releases were analyzed for fuel cooled for periods of 72, 240, and 1000 hours to determine the effects of decay cooling on these types of releases. Simulations were run with REM 123 Computer Code (Ref. A.2 Appendix E) with and without iodine filter credit. The results and details of the study are included in Appendix E, Figures E.2 and E.3. I The existing procedures and technical specification require that the containment integrity should be maintained during removal and replacement of the reactor vessel head and for movement of any heavy load over the core with the RV head removed. A complete containment integrity, however, is not required for movement of certain loads by the auxiliary hook (capacity I 10 tons) over the reactor vessel with the RV head in place. CEDM cooling duct / piping (Approx. Wt. 2000 lbs, Approx. Size 30'x3') and CEDM/In Core Detector Cable Trays (Approx. Wt. 3500 lbs, Approx. Size 20'x6') are the only identifiable I heavy loads which fall in this category. These loads are moved only twice during the refueling outage. In addition there are some loads such as ICI flanges, coupling tools, etc., I which also fall into this category but are not classified as heavy loads. I The movements of these loads are govered by written procedures. The loads are lifted no higher than necessary to safely clear the obstacles along the load paths. This minimizes the potential for a load drop directly over the reactor vessel. During these I operations even if any of the aforementioned loads are dropped, it is very unlikely that the drop will cause any significant damage to the RV head (6-3/16" thick steel). The kinetic energy I generated by the drop would essentially be absorbed in crushing of the shroud and ICI flanges etc. Two-blocking accidents resulting in RV head / core damage were considered highly improbable because of the following: I I I ' I
I I Main hook: The load block with empty hook is allowed to travel over the reactor vessel for the purpose of reversing the trolley direction. During this operation the hook is I empty and is not being moved up or down. Also the main hook is provided wth a Lever Actuated Limit Switch and a Geared Rotary Limit Switch. However one of the switches (Geared Rotary I Limit) is not wired for the upper limit. Subject to the wiring of this limit switch, the main hook will have redundant limit switches (see Section 4.0 Proposed Corrective Action No. 4). It I is therefore highly improbable that a two-blocking accident resulting in damage to the RV head or the core would occur. Auxiliary hook: As stated earlier, the auxiliary hook is I used for lifting of CEDM ducts and cable trays. During this lifting operation, the load block for the auxiliary hook is not positioned over the reactor vessel. Also the auxiliary hook I is provided wth a Lever Actuated Limit Switch and a Geared Rotary Limit Switch. However one of the switches (Geared Rotary Limit) is not wired for the upper limit. Subject to the wiring of this limit switch, the auxiliary hook will have redundant limit I switches (see Section 4.0 Proposed Corrective Actior No. 4). is therefore highly improbable that a two-blocking accident resulting in damage to the RV head or the core would occur. It Based on the above analysis it is concluded that because of the restrictions imposed on the violation of containment integrity and movement of heavy loads over the reactor I vessel, the OPPD Fort Calhoun refueling operations in Containment are in compliance with NUREG-0612 Criteria I Section 5.1 guidelines for a potential radiological release I due to the accidental drop of a heavy load from the Polar Crane onto a spent reactor core. Criteria II The load lifted by the Polar Crane that has the greatest potential for crunhing the refueled reactor core during refueling operations l is the Reactor Vessel Closure Head with the lifting rig attached. 5 Since this is considered to be the worst case scenario, all other heavy load drop accidents are encompassed by this analysis. The loads evaluated by this part of the analysis include the i Upper Guide Structure (UGS) and lifting Rig, as well as the concrete missile shields and lifting Rig. l The bounding worst case accident is postulated to be a fully-crushed core loaded with 133 new 4 weight-percent enriched fuel assemblies containing standard boron shim rods (FSAR Fig. 3.4-1) ,I with all control element assemblies inserted. The accident assumes a 36-foot drop of the RV Head and lift rig onto the UGS with all energy of the drop uniformly crushing the core in the l i I vertical direction. The RV Head and lift assembly are assumed to impact at an angle on the UGS, such that the maximum vertical crush is experienced. !I I I Specific initial conditions and assumptions used in the criticality analyses are documented in Appendix D, Table I D.I. The values of k which result from core crushing aresummarizedinTabibD.2. This analysis shows that k remains less than 0.95 on crushing and thus Criteria II oF Section 5.1 and Appendix A, Section 4.2.2 of NUREG 0612 are I satisfied. A detailed discussion of the KENO IV criticality code and other pertinent information is provided in Appendix D of this report. These criticality analyses indicate that the OPPD Fort Calhoun refueling operations in the Containment are in compliance with NUREG 0612 Criteria II Section 5.1 guideline I for criticality potential due to the accidental drop of a heavy load onto a refueled reactor core during refueling operations. Criteria III The load lifted by the Polar Crane that has the greatesc I energy potential for damage to the reactor vessel is the Reactor Vessel Closure Head with the closure head lifting rig assembly attached. The combined weight of 120 tons and I maximum lif t height of 36 feet exceed the drop energies of other lifted loads. Accordingly all other possible load combinations are encompassed by this load drop analysis of the Reactor Vessel Closure Head and Lifting Rig. The accident with the maximum energy for the Reactor Vessel Closure Head and Lifting Rig drop could occur when the head is I being replaced after refueling is complete. The cavity is drained of water to a level one foot t: low the rim of the reactor vessel. If a crane failure occurred when the load was at maximum height (36') directly over the reactor vessel, I there would be no fluid drag to reduce the energy at impact. It should be noted that, of 43 PWR's surveyed, assuming 1 I head lif t per year for refueling, more than 300 head lifts have been successfully completed without a head being dropped. This spans a period of time from 1961 to the present which is a total of 20 years without a single head drop incident. This indicates that the accident scenario described above is an unprobable occurrence. I To conservatively demonstrate compliance with NUREG 0612 Section 5.1 Criteria III, the worst possible damage to the Reactor Vessel was assumed. I I I I
From a loss of coolant standpoint, the worst damage that could occur due to the drop of the Reactor Vessel Closure Head and Lifting Rig would be the shearing of one or more of the reactor I vessel nozzles. Since the reactor core is located below the reactor vessel nozzles, the shearing of the nozzles would not immediately uncover the core. Even in the unlikely event of all I six RV nozzles being sheared off and the reactor vessel falling into the vessel cavity, (the reactor vessel support is at the nozzles) the vessel would remain essentially upright with the core covered by coolant. The heat transfer analysis indicates I that the heat removal for this reactor vessel configuration is adequate to keep the core cool. In the event this situation occurred, borated water could be provided by the reactor loops I or the Safety Injection Refueling Water tank discharging into the RV cavity to maintain core coverage. The structural analysis of the reactor vessel cavity indicates, that it will retain its integrity after the load drop. To ensure coolant is retained by the cavity, after the load drop, accesses to the cavity (Ref. GHDR Drawing No. 11405-A-14, Appendix B) will be assessed for leakage (see Section 4.0 Proposed Corrective Action No. 3). Procedures I for maintaining shutdown cooling after a load drop accident are located in Table A-1. I These analyses indicate that the OPPD Fort Calhoun refueling operations in the containment are in compliance with NUREG 0612, Criteria III of Section 5.1, guideline for damage to the reactor vessel following accidental dropping of a postulated heavy load. 2.4 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN PLANT AREAS CONTAINING EQUIPMENT REQUIRED FOR REACTOR SHUTDOWN, CORE DECAY
/ HEAT REMOVAL, OR SPENT FUEL POOL COOLING NUREG 0612, Section 5.1.5, provides guide 2ines concerning the design and operation of load-handling systems in the vicinity of equipment or I components required for safe reactor shutdown and decay heat removal.
Information provided in response to this section should be sufficient to demonstrate that adequate measures have been taken to ensure that in ll these areas, either the likelihood of a load drop which might prevent 3 safe reactor shutdown or prohibit continued decay heat removal is extremely small, or that damage to such equipment from load drops will be limited in order not to result in the loss of these safety-related iI functions. Cranes which must be evaluated in this section have been previously identified in your response to 2.1-1, and their loads in your response to 2.1-3c.
- 1. Ident.ify any cranes listed in 2.1-1, 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
,I carried and the basis for this evaluation (i.e., complete , compliance with NUREG 0612, Section 5.1.6, or partial compliance ! supplemented by suitable alternatives or additional design l features). For each crane so evaluated, provide the l load-handling-system (i.e., crane-load-combination) information specified in Attachment 1. 'I lI 1
I l RESPONSE: No exceptions. The main hook of the Auxiliary Building Crane is single failure proof, however, credit for this feature was not considerei during system analysis.
- 2. For any cranes identified in 2.1-1 not designated as single-failure proof in 2,4-1, a comprehensive hazard evaluation should be provided w.iich includes the following information:
- a. The presentation in a matrix format of all heavy loads and potential impact areas where damage might occur to Safety-related I equipment. Heavy loads identification should include designation and weight or cross-reference to information provided in 2.1-3-c.
Inpact areas should be identified by construction zones and eleva-tions or by some other method such that the impact area can be located on the plant general arrangement drawings. Figure 1 provides a typical matrix. RESPONSE: Tables A-1 through A-7 and the cross-referenced figures in Appendix B respond to this question in a matrix format. I b. For each interaction identified, indicate which of the load and impact area combinations can be eliminated because of separation and redundancy of safety-related equipment, I mechanical stops and/or electrical interlocks, or other site-specific consideration. Elimination on the basis of the aforementioned considerations should be supplemented by the , following specific information: (1) For load / target combinations eliminated because of separation and redundancy of safety-related equipment, I discuss the basis for determining that load drops will not affece continued system operation (i.e., the ability of the system to perform its safety-related function). (2) Where mechanical stops or electrical interlocks are to be provided, present details showing the areas where , crane travel will be prchibited. Additionally, provide ll '5 a discussion concerning the procedures that are to be used for authorizing the bypassing of interlocks or removable stops, for verifying that interlocks are functional prior to crane use, and for verifying that interlocks are restored to operability after operations which require bypassing have been completed. (3) Where load / target combinations are eliminated on the basis of other, site-specific considerations (e.g., l maintenance sequencing), provide present and/or proposed
- B3 technical specifications and discuss administrative procedures or physical constraints invoked to ensure the l continued validity of such considerations.
l lI I
t l l I _ t i . i I RESPONSE: See response to ite4[22a above ' I c. For interactions not eliminated by the analysis of 2.4-2-b, above, identity any hacdling systems for sp9cific loads which you have evaluated as having sufficient design features to make the liklihood of a load drop extremely small and the basis I for this evaluation (e.e., completS compliance with NUREG 0612, Section 5.1.6, or partial complisoce supplemented by suitable alternative or additional design features). For each crane so evaluated, provide the load-handling-systen.(1.e., crane-load-combination) informatien specified in Attachment 1. RESPONSE: Not Applicable i
- d. -For interactions not eliminated in 2.4-2-b or 2.4-2-c, above, demonstrate.using appropriate analysis that damage would not preclude operation of sufficient equipment to allow the system to perform its safety function following a load drop (NUREG 0612, Section 5.1, Criterion IV). For each analysis so conducted, the following information should be provided:
(1) An indication of whether or not, for the specific load being investigated, the overhead crane-handling system I is designed and constructed such that the hoisting system vill retain its load in the event of seismic accelerations equivalent to those of a safe shutdown earthquake (SSE). I RESPONSE: The only crane analyzed to demonstrate load retention capabilities ~ during an SSE sis the main book of 6.he I - Auxiliary Building Crane, which is single failure proof and included in 2.4.1 above.(see Appendix C, Attachment 1 for documentation). (2) The basis for any exceptions taken to the analytical guidelines of NUREG 0612, Appendix A. RESPONSE: No exceptions (3) The information requested in Attachment 4. RESPONSE: Not applicable I __ ____________----- -- --- J
LO AD / DROP TABLE - A-I I of 3 CRANE CONTATNNijNT POLAR CRANE LOCATION CONTAlurlENT HEAVY LOAD RV llEAD & LIFT RIG 120 T, LOWER LOAD IMPACT AREA SEC FIGURES IN APPENDIX B AND FIGURE B-1 ItLOCK ST (MOVED DURING REFUELING OUTAGE ONLY) ELEVATION SAFETY-RELATED HAZARD ELIMINATION STATEME N T-EQUIPMENT , (1) fia4"-0" *(l) SilUTDOWN C001.tNG (1) CASE (I) EL. 994' PIPINt; IF A LOAD DROP SilEARS OFF PIPING BETWEEN Tile CONTAINMENT PENETRATIONS AND VALVES IICV-327, IICV-329, llCV-331,llCV-333, AND llCV-348, TIIESE VALVES AND Tile VALVES OUT-SIDE Gi CONTAINHENT CAN BE CI.0 SED TO STOP DRAINING OF Tile SYSTEtt OR Tile REACTOR CAVITY. Tile TECII. SPECS ALLOW 8 Il0URS FOR TIIE SYSTEM TO IlE INOPERAI11.E IF NOT IN Rt.iUELING. TilEREFORE, Tile PLANT COULD WEI.D A BLIND FLANCE ON Tile NECESSARY PIPING ENDS AND PLACE A PORTION OF Tile C00LDOWN SYSTEM BACK IN OPERATION III Tile 8 Il0URS TIME LIMIT. . .NO IIAZARD FOR CASE (1) . CASE (II) EL. 1013' IF LUAD DROP SilEARS VALVE HCV-348,. SYSTEM LOW POINT AT E. 1003', Tile REACTOR CAVITY COULD DRAIN, WITil NO OPERATOR ACTION, 'IO Tile BOTTOM OF Tile Il0T I.EG PIPF. AT AN APPROX. EL. OF 1005'. TilIS WOULD LEAVE APPROX. 4' 0F WATER COVEl,11NG Tile CORE. IT S110ULD BE NOTED,110 WEVER, TilAT Tile LOWER CAVITY AREA WII.T. NOT DRAIN bel.0W 1013' 11UE TO THE CONCRETE WALL SEPARATING Tile VESSEL FROH Tile I.0WER CAVITY. Tile TECil SPECS ALLOW Tile SYSTEM TO BE INOPERAlli,E FOR A IfAXItfUM OF 8 Il0llRS AND Tl!E Pl. ANT STAFF COUI.D REPAIR Tile BREAK WITilIN TilAT TIlfE PERIOD. . .NO IIAZARI) FOlt CASE (II). (2) 994'-0" *(2) SAFETY-INJECTION (2) IF A I,0AD DROPS IN Till: AREA BOUND BY COLUMN LINES 10 & 11, Tile CONTAINMENT WALI, AND CIIARGING PIPJNG AND STEAM GENERA'10R RC-211 WALL, IT COULD SilEAR Tile IIPSI, LPSI, AND CIIARGING PIPING. ,TilIS WOULD RESULT IN Tile ELIMINATION OF ALL NORtlAL PATilS TO IN. LECT WATER INTO Tile REACTOR COOLANT SYSTEM. SINCE TilERE IS NO PROCEDURE TO IN.1Ecr WATER IN TIIE RC SYSTEM FOR TIIIS CASE, TilIS 1.0AD DROP APPEARS TO BE A POSSIBLE IIAZARD, (SEE CONTINilATION ON NEXT PAGE).
- EQUIPt!ENT REFERENCED TO TilIS NOTE IS Tile ONI.Y SAFETY RELATED EQUIPMENT ItEQIllRED To llEET NUREG-0612 CRITERIA WilICil IS LOCA'TED WITilIN Tile LOAD DROP ENVEI,0PE.
M ' E M M M M M M M M M M M
LO AD / DROP TABLE - A-l 2 or 3 CRANE CONTAINtIENT POLAR CRANE (CONT.) LOCATION CONTAINMENT HEAVY LOAD RV HEAD & I,IFT RIG 120T. LOWER LOAD BLOCK IMPACT AREA SEE FIGURES IN APPENDIX B AND FIGURE B-1 ST ELEVATION SAFETY-RELATED HAZARD ELIMINATION STATEMENT EQUIPMENT - (2) 994'-0" *(2) SAFl?rY-IN.3ECTION (2) CONTINUED CONTINUED CIIARGING PIPING IF AI.L NORifAL WATER INJECTION PATilS ARE ELIMillATED, AN ALTERNATE 8'ATil CAN BE MADE AVAILABLE BY USING TIIE FOLLOWING LINE-UP: ALLOW WATER FR0ff TIIE BROKEN PIPES TO DRAIN INTO TIIE RECIRCULATION SUMP AREA. ESTABLIS11 A FLOW PATil So WATER DISCIIARGED FROM Tile LPSI OR CONTAINMENT SPRAY PUMP IS ROUTED TilROUGil Tile Sl!UTDOWN COOLING llEAT EXCIIANGER. OPEN i0CKED CLOSE VALVES SI-342 AND SI-185, ESTABLISil A FLOW PATil TO TIIE REACTOR LOOPS VIA' AN ACCUl!UI.ATOR. USE Tile LPSI OR CONTAINMENT SPRAY PUMP TO MOVE WATER TIIROUGil TIIE CORE TAKING SUCTION ON Tile RECURCULATION SUMP UNTIL Tile SilUTDOWN COOLING PIPING CAN BE REPAIRED. IN Tile IINLIKELY EVENT TIIAT TIIE BRANCII LEG OF Tile ACCUllULATOR FII,I. PIPING WERE TO BE SilEARED WITil Tile SilUTDOWN COOLING PIPING, TIIIS TWO INCII BRANCll LEG COULD BE PLUGGED ALLOWING Tile ALTERNATE SYSTEM TO COOL Tile CORE UNTII. Tile SillrIDOWN COOLING SYSTEM IS REPAIRED. IF Tile RECIRCULATION SUllP STRAINER WERE TO BECOME CLOCCED BECAUSE OF ITEllS USED DURING Tile OUTAGE, AN AIR 110SE WOULD BE CONNECTED TO Tile TEST CON!!ECTIONS UPSTREAM OF Tile CilECK VALVES SI-160 & SI-159 AND AIR USED TO KEEP Tile STRAINERS CLEAR WilILE TilIS ALTERNATE SYSTEM IS OPERATING ALLOWING REPAIR OF Tile NORMAL COOLING PIPING TO COMMENCE. TilIS WRITTEN PROCEDURE WILL EI,IMINATE AN1f IIAZARDS DURING OPERATION FROM Tile CRANE IN TIIOSE AREAS OF CONCERN 10 MAINTAIN SAFE SilUTDOWN. (SEE SEC. 4.0 PROPOSED CORRECTIVE ACTION NO. 1) (1) 1013'-0" *(3) REACTOR VESSEI, (3) IN Tile UNLIKELY EVENT OF Tile REACTOR VESSEL llEAD AND LIFT RIG FALI,ING ON Tile REACTOR VESSEL, Tile WORST CASE CONSEQUENCES WOULD BE Tile SilEARING OF
, ALL Tile N0ZZLES W111Cil SUPPORT Tile REACTOR V8?SSEI.. Tile VESSEL WOUI,D TilEN DROP INTO Tile REACTOR VESSEL CAVITY.
Tile CAVITY WOULD FII.L WITil WATER FROM TIIE BROKEN LOCP N0ZZLES. ADDITIONAL WATER WOULD BE SUPPLIED FROM Tile ACCUt!ULATORS AND SIRW TANK IF REQUIRED. FLOW WOULD BE ESTABLISIIED TIIROUCil Tile SilUTDOWN COOLING SYSTEll AND INTO Tile
- EQUIPMENT REFERENCED TO TIIIS NOTE IS Tile CNI,Y SAFETY REALTED EQUIPMENT REQUIRl;D TO MEET NUREG-0612 CRITERIA WilICII IS I,0CATED WITilIN TIIE LOAD DROP ENVEl.0PE.
W W m a m W m m A m W W m W m m
LO AD / DROP TABLE - A-1 3 of 3 CRANE ._coNTATMfFNT POLARCRANR (cnNT a LOCATION cnuratuurnt HEAVY LOAD.av_linAn F. LIFT RIG 120T. LOWER LOAD ELOCK IMPACT AREA _SEE_UGURES_ULAPPRNnTY B AND FIGURR R-1 ST ELEVATION SAFETY-RELATED-EQUIPMENT. HAZARD ELIMINATION STATEMENT. . (3) 10.13'-0" (3) REACTOR VESSEL (3) CONTINUED CONTIiWCD CAVITY VIA Tile BROKEN N0f2LES. ' WATER WOULD TilEN RETURN TO Til.E SilUT DOWN COOLING SYSTEM VIA Tile BROKEN N0ZZLE ON Tile RETURN PIPING LOOP. Tile IIEAT REMOVAL CALCULATION S110WS TilAT Tile HEAT REMOVAL FOR TilIS REACTOR VESSEL CONFIGURATION IS ADEQUATE TO KEEP TI:E CORE COOL. TilEREFORE, NO IIAZARD EXISTS. .
* . EQUIP!!ENT REFERENCED TO TIIIS NOTE IS Til'E ONLY SAFETY RELATED hQllIPSIENT REQUIRED TO llEET l'lIREG 0612 CRITERIA WilICII IS LOCATED WITilIN Tile LOAD DROP ENVELOPE.
W m M M m M M M M M M M M M M M M M M
LO AD /DR OP TABLE - A-2 I or i CRANE AtixlLI ARY BUILDIllG CRANE (BRIDGE) LOCATION AUXILI ARY BUII. DING HEAVY LOAD 5.$ TON SOLID WASTE SilIPIING CONTAltIER IMPACT AREA SEE ITGURES IN APPENDIX B AND FIGURE B-2 ELEVATION SAFETY-RELATED- HAZARD ELIMl NATION STATEMENT, - EQUIPMENT (1) 1007'-0" *(l) SAFETY IN.IECTION (1) IF A 1.0AD IS DROPPED INTO Tile S.I.R.W. TANK TIIE PLAllT CAN STII.L ACIIIEVE A SAFE REFUELING WATER SilUTDOWN CONDITION AND NO IIAZARD EXISTS. TANK e
- EQtlIPMENT. REl ERENCED TO TilIS NOTE IS Tile ONI.Y SAFETY RELATED EQUIPMENT REQUIRED TO ffEET NUREG-06'12 CRITERIA WIIICil IS I,0CATED WITIIIN Tile I0AD DROP ENVEIOPE.
W W um W m M I e W m W W W M M M M M M
LO AD / DROP TABLE-A-1 ] of I CRANE INTAKE STRUCTURE CRANE (BRIDGE) LOCATION INTAKE STRUCTURE HEAVY LO.' 3 CIRCULATING WATER PUMP FI370F (24,000 LB.) IMPACT AREA { SEE FIGURES IN APPENDIX B AND Ei_EVATION SAFETY-RELATED HAZARD ELIMINATION STATEMENT ; - EOUIPMENT. (1) 993'-6" *(1) RAW WATER PUMPS AND (1) Tile RAW WATER SYSTEM IS NOT REQUIRED FOR SAFE SilUTDOWN. Il0 WEVER, ONLY LIMITED ASSOCIATED PIPING RAW WATER OUTAGE CAN BE ALLOWED DUkING REFUELING OUTACE. IF Tile CIRCULATING WATER PUMP MOTOR WERE DROPPED' AND CRUSilED RAW WATER (RW) PUMP AC-10A OR 10D AND ASSOCIATED PIPING, Tile RW PUMPS.10A AND 10B WOULD SUPPLY WATER TO COOL
. COMPONENTS BY CLOSING TIIE APPROPRIATE CROSS-TIE VALVES.. IF Tile LOAD WERE DROPPED ON RW PUMPS AC-10B & 10C, Tile CROSS-TIE VALVES COULD BE' CLOSED AND COMPONENTS COULD BE COOLED BY AC-10A OR 10D. -
(2) 993'-6" *(2) ELECTRICAL MOTOR (2) IF A LOAD IS DROPPED NEAR Tile SOUTil WALL, IT COULD SilEAR ALL RAW WATER PUMP LEADS TO Tile RAW CABLES. A PLANT PROCEDURE WILL BE WRITTEN TO ENSURE TilAT Tile LOSS OF Tile RAW WATER PUMPS WATER PUMP CABLES DOES NOT IN1!IBIT DECAY llEAT REMOVAL CAPABILITY DURING RE-FUELING OUTAGE. , (SEE SEC. 4.0 PROPOSED CORRECTIVE ACTION NO. 2) f t-(3) 993'-6" *(3) RAW WATER AREA (3) IF Tile. LOAD IS DROPPED TilRU Tile RAW WATER FLOOR SLAB WilEN Tile RIVER IS AT FLOOR SLAB NORMAL LEVEL OF 992'-0", WATER WOULD NOT ENTER Tile CUBICLE. IF Tile 1 RIVER LEVEL WERE AT 998'-0", (A 1% FLOOD STAGE), AND Tile LOAD WENT TilR0llGli Tile RAW WATER FLOOR SLAB, Tile SLUICE GATES WOULD BE CLOSED AND ANY WATER IN Tile PUMP
' CUBICLE PUMPED OUT WITil A SUBMERSIBLE PUMP. TilEREFORE, No llAZARD EXISTS.
i i
*' EQUIPl!ENT REFERENCED TO T111S NOTE IS Til'F, ONLY SAFETY llELATEt) tQlIIPilENT j
t REQUIRED To HEET NUREG 0612 CRITERIA WillCil IS LOCATED WITilIN Tile LOAD DROP ENVELOPE. M M EB M 955 M M M M M WiD W W mmmmem
LO AD / DROP TABLE - A-4 _I__ of I 4 CRANE _concRETF. SLAB REMOVAL CRANE #1 010NORAIL) LOCATION AUXILIARY BUII.TDNG HEAVY LOAD RE110VABLE CONCRETE SLAB 16,400 LB. IMPAC T AREA SEE FIGURES IN APPF*~'IX B AND FICURE B-4 ELEVATION SAFETY-RELATED HAZARD ELIMINATION STATEMENT.. EQUIPMENT (1) 984'-4" (1) CABLE # EB-9372-FEEDS (1) REDUNDANCY IS SUPPLIED BY SI-1 A WilICil RECEIVES POWER FROM " TRAIN A" IICV-2937 SI-1B SUCTION AND IS NOT AFFECTED BY TilIS LOAD DROP.
- VALVE (" TRAIN B")
(2) 984'-4" (2) CABLE # EB-9373-FEEDS (2) REDUNDANCY IS SUPPLIED BY SI-1A WilICll RECEIVES POWER FROM " TRAIN A" IICV-2938 SI-IB 'AND IS NOT AFFECTED BY TilIS LOAD DROP. IN ADDITION, llCV-2937 T. IICV-2938 DISCIIARGE VALVE ARE NORMALLY OPEN. LOSS OF POWER TO Tile VALVES WILL MAINTAIN TilDI IN Tile (" TRAIN B") OPEN POSITION.
- EQUIPl!ENT REFERENCED TO TiliS NOTE IS TIIE ONLY SAFETY RELATED EQUIPilENT REQUIRED TO MEET NUREG 0612 CRITERIA WilICll IS LOCATED WITilIN Tile LOAD
, DROP ENVELOPE. -
E W EE W M M ES O E E E E E E O E E Y E
LO AD / DROP TABLE - A-5 I of 2 CRANE CONCRETE SLAB REll0 VAL CRANE #2 (HONORAIL) LOCATION AUXILIARY BUII. DING HEAVY LOAD RE}10VAllLE CONCRETE SLAB 12,5'00 LB. IMPACT AREA SEE FIGURES IN APPENDIX B AND FIGURE B-5 ELEVATION SAFETY-RELATED HAZARD ELIMINATION STATEMENT: EQUIPMENT - - (1) 989'-0" *(l) 2" CilARGING LINE & (1) TilESE LINES ARE NOT REQUIRED,TO ACllEIVE A SAFE SilUTDOWN CONDITION BECAllSE ALTER-2" AI. TERNATE CllARG- NATE WATER CAN BE SUPPLIED BY USE OF Tile IIPSI PUtiPS. TilEREFORE NO IIA 7.ARD EXISTS. ING LINE FROH CilARG-ING PUMPS TO Tile REACTOR C001. ANT LOOPS (2) 971'-0" *(2) 12" SilUTDOWN COOL- (2) IF SilEARED DURING NORilAL OPERATI0ll, CO TO IlOT S11UTDOWN AND REPAIR. IF SilEARED ING SUCTION AND DURING REFilELING, TECil. SPEC. STATES TiiAT TILE SYSTEM CAN DE INOPERABl.E I 0R 8 DISCllARGE PIPING 110URS, SO TilAT A NEW SPOOL PIECE CAtt BE WELDED IN TO CONTINUE PROVIDING COOLING WATER. (3) 971'0 " *(3) 'B' CABLE TRAY (3) REDUNDANCY SUPPLIED BY llPSI PtRIP SI-2A LPSI PUMP SI-IA WilICll ARE FED Filott " TRAIN SUPPI.Y To llPSI PIRIP A" POWER. Tile " TRAIN A" CABLE TRAYS ARE NOT AFFECTED BY TIIIS LOAD 11 ROP. SI-2B AND LPSI PUMP SI-1B (4) 989'-0" *(4) EI.ECTRICAL SUPPLY TO :(4) REDUNDANCY IS ACllEIVED BY ALTERNATE PUMP (AC-5A 011 AC-5B DEPENDING ON Tile T.0SS SPENT FUEL STORAGE OF ONE PUllP). TilESE PUllPS ARE FED FROH AI, TERNATE POWER SOURCES AND ONIY ONE PUMP AC-5A OR SB, PUMP WOULD BE LOST AT ONE TIME. CilARGING PUMP Cll-lB PACKING COOLING - PUMP Cll-IC, BORIC ACID PUHP CII-4B, I. PSI SI-1B, I. PSI SI-l B 110 TOR LEADS. BORIC ACID GRAVITY FEED VALVE IICV-258, SilUTDOWN COOLING VALVE IICV-347, Rff-061 & Ril-062 M M SE M M W IN W W W W W W M M M M M M
LO AD / DROP TABL E - A-5 2 d 2 CRANE CONCRETE SI.AB RE!!OVAT. CRANE #2 (HONORATI.) LOCATION AUXII.I ARY BUTI. DING HEAVY LOAD RE!!OVABI.E CONCRETE St.AB 12.500 1.B. IMPACT AREA SEE FIGURES IN APPENDIX B AND FIGURE B-5 ELEVATION SAFE TY- REL ATED HAZARD ELIMINATION STATEMENT EQUIPMENT _. (5) 989'-0" *(5) El.ECTRICAI SUPPL.Y To (5) REDUNDANCY SUPPI.IED BY " DIVISION B" POWER SUPPI.Y TO Cll-1B AND I'ACKING C001.It G PACKING COOLING PUt!P PUllP IB. CII-1A, CilARGING PUllP . Cll-1A
- EQUIPl!ENT REFERENCED TO TilIS NOTE IS Tile ONI.Y SAFETY RELATED I?QtlIPflENT REQtilRI'.D TO IIEET NUREG-0612 CRITERIA WilICII TS LOCATED WITilIH Tile I.0AD DROP ENVEI.0PE.
M M 'Ml M M M GY M M M Y' M & M M M & M M
LO AD / DROP TABLE - A-6 I of I CRANE WASTE EVAPORATOR EQUIPMENT CRANE (MONORAIL) LOCATION AUXILIARY Bt!ILDING HEAVY LOAD RElf0VABI.E CONCRETE SLAB 19,800 LB. SEE FIGURES IN APPENDIX B AND' FIGURE B-6 IMPACT AR.EA ELEVATION SAFETY-RELATED HAZARD ELIMINATION STATEMENT, . EQUIPMENT (1) 989'-0" *(l) 2" CIIARGING PUMP (1) TIIIS PIPE IS NOT REQUIRED TO Acil EVE SAFE SilllTDOWN BECAUSE ALTERNATE WATER RETURN PIPING To CAN BE SUPPLIED BY USE OF Tile IIPSI PUMPS. TilERLPORE. UO IIAZARD EXISTS. REACTOR COOLANT LOOPS (2) 1007'-0" *(2) 4" RAW WATER SUPPLY (2) IF PIPES WERE SilEARED DURING REFUELING MODE, Tile LPSI SEAL DESICH IS RATED AND RETURN PIPING FOR 300 F OPERATION. TilIS WILL ALLOW Tile LPSI PUMPS TO OPERATE WITil0UT SEAL FOR S.I. AND CON- COOLING. TAINMENT SPRAY COOLERS (3) 989'-0" *(3) 8" SUPPL.Y AND RE- (3) IF SilEARED DURING NORMAL OPERATION, GO TO Il0T SilVTDOWN, USE Tile SPENT FUEL TURN PIPING FOR Tile POOL CROSS-TIE TO COOL Tile SPENT FUEL POOL. IF SilEARED IN A REFUELING MODE, SPENT FUEL POOL INSTAL,L Ti!E SPENT FUEL POOL CROSS-TIE AND PROVIDE COOLING TO Tile P001. AND IIEAT EXCIIANGER REACTOR PIPING IN ACCORDANCE WITil Tile TECil. SPEC. (4) 971'-0" *(4) 10" RETURN PIPING (4) IF SilEARED IN NORMAL OPERATION, GO TO Il0T SIIUTDOWN AND FIX. IF SilEARED IN RE ' FOR Tile SilUTDOWN FUELIMG MODE, ISOLATE, CUT OFF PIPE END, AND WELD BLIND FLANGE TO PIPE. TECll. COOLING SYSTElf SPEC. STATES SYSTEM CAN BE OUT OF SERVICE FOR 8 Il0URS. (5) 971'-0" *(5) LPSI PUMP SI-1B (5) IF CRUSilED, A REDUNDANT LPSI PUllP SI-1A IS AVAILABLE TO PROVIDE SAFE SilUTDOWN.
- (6) 989'-0" *(6) CIIARGING PUMP Cll-lC (6) IF CRUSilED, A REDUNDANT CilARGING PUMP IS SUPPLIED FOR SAFE SIIUTDOWN.
I ! (7) 997'-0" *(7) LPSI PUl!PS SI-1B & (7) Tills CREATEF A LOSS OF ALL CllARGING PUMPS OR I.0SS OF BOTil LPSI PUMPS. TilIS IS l SI-I A !!OTOR I.EADS, NOT A IIAZARD BECAUSE POWER CAN BE SUPPLIED TO ALTERNATE PUMPS AND STILI, PROVIDE ! LPSI PUMP SI-1B WATER TO Tile REACTOR COOLANT SYSTEM. I SUCTION AND DIS-CIIARGE VALVES POWER , CABLE, CllARGING PulfPS Cll-1 A, CII-1B, Cll-lC ELECTRIC LEADS
- EQUIPMENT REFERENCED TO Tills NOTE IS Tile ONI,Y SAFETY RELATED EQUIPHENT REQllIRED E W jE M MM M
- M ?ET M ?G- M CR W A W I I W AT M ITli W lE E Dl W NVI M . M
LO AD / DROP TABLE - A-7 1 of I CRANE DEBORATING DEHINERALIZER CRANE (MONORAIL) LOCATION AUXILIARY BUILDH;G HEAVY LOAD REMOVABLE CONCRETE SLAB 12,600 LB. IMPACT AREA SEE FIGURES IN APPENDIX B AND FIGURE B-7 ELEVATION SAFETY-RELATED EQUIPMENT HAZARD ELIMINATION STATEMENT. r (1) 989'-0" *(I) STORAGE POOL llEAT (1) IF IIEAT EXCIIAN'FR AND PUMP AC-5A 'ARE CRUSilED IN NORMAL OPERATING !! ODE, CO TO EXCIIANGER AND SPEN 1 Il0T SIIUTDOWN, IhsTALL TIIE SPENT FUEL POOL CROSS-TIE AND COOL Tile POOL USING FUEL POOL PUMP LPSI PUMPS. IF ABOVE ITEMS ARE CURSilED IN REFUELING MODE, INSTALL Tile SPENT AC-5A FUEL POOL CROSS-TIE AND PROVIDE COOLING TO Tile POOL AND RCACTOR PIPING IN ACCORDANCE WITil Tile TECll. SPEC. (2) 971'-0" & *(2) ELECTRICAL SUPPLY (2) REDUNDANCY SUPPLIED BY PIIYSICAL SEPARATION OF COMPONENTS AND CABLE TRAYS AND 989'-0" TO Tile FOLLOWING: WOULD PROVIDE SAFE SilUTDOWN. LPSI PUHP SI-1B SUCTION AND DIS-CllARGE VAI.VES, CIIARGING PUMP CII-1B AND Cll-lC 3 STORAGE POOL PUMP , AC-5A, BORIC ACID CRAVITY FEED VALVE IICV-258 t EQUIPMENT REFERENCED TO TilIS NOTE IC Tile ONLY SAFETY RELATED EQUIP!!ENT REQUIRED TO MEET NUREG-0612 CRITERIA WilICll IS I.0CATED WITilIN Tile LOAD DROP ENVEl.0PF M M $l M W W EE E E E O' E N U N' O N' O E
I 4.0 PROPOSED CORRECTIVE ACTIONS I The following corrective actions are planned based upon the results of the design evaluation and analysis of Control of Heavy Loads at Nuclear Power Plants.
- 1. A procedure will be written to provide an alternate path for I shutdown cooling water in the event of a load drop in the area bounded by Columns 10 and 11 and the biological shield wall in the contsinment (Ref. GHDR #11405-A-5, Appendix B). This procedure will permit the use of the containment polar crane in the area.
I 2. (See Load / Drop Table A-1) A procedure will be written to prevent the loss of the raw water I pumps due to a load drop accident destroying the power supply cables. The procedure will: I a) Prohibit loads from being carried over the area above the cable tray supplying power to all four raw water pumps, and/or b) Outlire emergency repair procedures to connect the fire pump I . discharge into the Raw water header to provide component cooling during the repair of the Raw water pump power cables. I 3. The design of the access door to the reactor vessel cavity at EL. 976'-0" (Ref. GHDR Drawing No. 11450-A-14, Appendix B) will be reviewed. This design evaluation will ensure that the door can I withstand hydrostatic pressure of -the flooded cavity after a postulated load drop shears off all the nozzles of the reactor vessel and the vessel falls into the cavity. If the door design is found to be deficient, appropriate steps will be taken to ensure the reactor core remains covered with coolant.
- 4. The Gearej Rotary Limit Switches will be wired for the upper limit on che main hack and the auxiliary hook of the containment I_ polar crane. This will provide redundent limit switches and prevent' a two-blocking accident.
5.0 SCHEDULE All proposed corrective actions (with the exception of Item 3) will be ,I implemented within two years of this submittal. Proposed Corrective Activu, Item 3 will be completed by the end of 1984 refueling outage. I I I . _ A
. _ _ . _ __ _= .. . . _ _ . _ . _
I l lI l3 ) i l l g iressoix i DECISION CHART I I !I
- 4 lI lE lI I
4 iI B lI 4 !,I _ . - __.--_. - - _ _ _
FIGURE A-1 PAGE 1 DECISION CHART NUREG 0612 I IDENTIFY CRANE , I . IDENTIFY LOADS LIFTED BY CRANE l , IDENTIFY AREA 0F CRANE TRAVEL IGNORING 1 l'ECHANICAL-ELECTRICAL INTERLOCKS g - IS YES ' - SINGLE-FAILURE PROOF? -
.1 U
NO ACCEPTABLE CONDITION ARE ALL LOADS I LIFTED BY CRANE LESS THAN THE WEIGHT OF A SINGLE FUEL BUNDLE = AND ITS HANDLING TOOL? I
/
NO 9 . SEE PAGES 2 THROUGH 5 I a
I FROM PAGE 2 PAGE 1 I CAN A LOAD DROPPED BY NO I THE CRANE LAND OR FALL IN r ACCEPTABLE CONDITION THE SPENT FUEL / " POO g YES U I ~ IS I RANE TRAVEL LIMIT BY INTERLOCKS TO PREVENT HOOK TRAVEL WITHIN 15' 0F THE SPENT FUEL I POOL OR FAR ENOUGH AWAY TO PREVENT A DROPPED LOAD (BY ANY FAILURE MODE) FROM FALLING y YES IN THE SPENT FUEL POOL? (NUREG 0612 SEC. 5.1.2-2) I NO - I = = I , IS . DOES I CRANE TRAVEL LIMITED BY INTERLOCKS TO PREVENT HOOK TRAVEL WITHIN 25' 0F YES , YES ANALYSIS INDICATE COMPLIANCE WITH CRITERIA I, II, AND III I " HOT" SPENT FUEL? (NUREG 0612 SEC.
.1.2-3 3
OF NUREG 06127 1 ACCEPTABLE CONDITION NO NO V o
=
9 u SUGGEST POSSIBLE i SOLUTIONS
l PAGE 3 FROM 1 PAGE I i I 1 CAN i A LOAD DROPPED NO T BY THE CRANE LAND ON THE r ACCEPTABLE CONDITION REACTOR VESSEL? 3 1 I , CAN A DROPPED CAN RAPID 'I " : LOAD IMPACT SPENT FUEL CAUSING A RADIO-ACTIVE RELEASE IN EXCESS OF r CONTAINMENT ISO-LATION BE USED TO PREVENT RADIOACTIVE YES ,g is THE LIMITS OF 10 RELEASE (NUREG 0612 5 CFR PART 100? SEC. 5.1.3-2) .I s I g" ACCEPTABLE CONDITION u NO NO - 1
'I CAN A l
l I q x DROPPED LOAD IMPACT SPENT FUEL CAUSING A K,gg> 0.957 YES F SUGGEST POSSIBLE SOLUTIONS l
~
n I " l I CAN A CAN A DROPPED LOAD SOURCE OF CAUSE A LOSS OF COOLANT BE PROVIDED COOLANT EXPOSING = TO OVERCOME LOSSES? SPENT FUEL?
-o I
YES 00 r ACCEPTABLE CONDITION
PAGE 4 FROM PAGE 1 I P CAN A I LOAD DROPPED , BY THE CRANE FALL ON ' 0 ' EQUIPF.NT, PIPING, WIRING, ETC. , ACCEPTABLE CONDITION I REQUIRED TO MAINTAIN SAFE SHUTDOWN? q YES I CAN SYSTEM
=
REDUNDANCY MAINTAIN i SAFE SHUTDOWN?
- I NO E
g OTHER SITE CON- l DITIONS AT THE TIld2 0F l l \ YES 1 LOAD DROP PERMIT SAFE SHUTDOWN ) TO BE MAINTAINED? 4 l I N ' I " NO IS l COMPONENT PRO-r SUGGEST POSSIBLE TECTED FROM DROPPED LOAD BY PLANT SOLUTIONS TRUCTUR h j I YES ANALYZE STRUCTURE PEh JURVIVES STRUCTURAL DESIGN e I CRITERIA W ACCEPTABLE CONDITION
PAGE 5 FROM I s PAGE 1 I P CAN A I LOAD DROPPED BY THE CRANE FALL ON EQUIPMENT, PIPING, UIRING, ETC., r ACCEPTABLE CONDITION REQUIRED TO ATTAIN SAFE SHUTDOWN? YES CAN SYSTEM I REDUNDANCY ATTAIN SAFE SHUTDOWN? I NO i I ARE LOADS CAPRIED OVER COMPO- -
" NENT ONLY WHEN PLANT IS IN I SAFE SHUTDOWN?
i i
- I a IS
~
no SUGGEST POSSIBLE TECTED FROM DROPPED LOAD 2
- BY PLA'IT SOLUTIONS TRUCTUR
- r k
ANAL s SURVIVES S RU M E M FAILS STRUCTURAL DESIGN (j CRITERIA E ,
, AcCmme C,Nem.
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I I I APPENDIX C INFORMATION REQUESTED IN ATTACHMENT 1, " SINGLE FAILURE PROOF HANDLING SYSTEMS" I l l' l I I l I I 'I I ,I I
I APPENDIX C SINGLE FAILURE PROOF HANDLING SYSTEMS
- 1. Provide the name of the manufacturer and the design-rated load (DRL).
I If the maximum critical load (MCL), as defined in NUREG 0554, is not the same as the DRL, provide this capacity. The Auxiliary Building Crane consists of Harnishfeger bridges I
RESPONSE
and end trucks with an Ederer Trolley and hoists. The design rated load (DRL) and maxiun2m critical load (MRL) are 75 tons for the main hook and 10 tons for the auxiliary hook. I 2. Provide a detailed evaluation of the overhead handling system with respect to the features of design, fabrication, inspection, testing, _nd operation as delineated in NUREG 0554 and supplemented by the identified alternatives specified in NUREG 0612, Appendix.C. This evaluation must include a point-by point comparison for each section of l NUREG 0554. If the alternatives of UUREG 0612, Appendix C, are.used for certain applications in lieu of complying with the recommendation of NUREG 0554, this should be explicitly stated. If an alternative to any of those contained in NUREG 0554 or NUREG 0612, Appendix C, is I proposed, details must be provided on the proposed alternative to demonstrate its equivalency.
' RESPONSE: Attachment I to Appendix C provides the requested information.
- 3. With respect to the seismic analysis employed to demonstrate that the overhead handling system can retain the load during a seismic event equal to a safe shutdown earthquake, provide a description of the ,
method of analysis, the assumptions used, and the mathematical model evaluated in the analysis. The description of assumptions should include the basis for selection of trolley and load position. RESPONSE: See above response
- 4. Provide an evaluation of the lifting devices for each single-failure-proof handling system with respect to the guidelines of NUREG 0612, Section 5.1.6.
RESPONSE: The only items lifted by the single failure proof main book will be the spent fuel shipping cask and sample cask. Neither cask has been selected at this time. However, it is anticipated that each will be provided with rigging and - attach points that meet or exceed the requirements of NUREG 0612.
- 5. Provide an evaluation of the interfacing lift points with respect to the guidelines of NUREG 0612, Section 5.1.6.
RESPONSE: See above response I C-1 I
~
I .
i 1 i 1 1 i a i ; 4 1 1 i I t t , i i 1 I I l a l ATIACHMENT 1 4 j TO I l i APPENDIX C O !k l l i I i r i l . l l l i I I e
= * =
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I I . SAFETY ANALYSIS REPORT I ON THE REFUELING AREA CRANE FOR HANDLING THE SPENT FUEL SHIPPING CASK OMAHA PUBLIC POWER DISTRICT FORT CALHOUN POVIER STATION, UNIT 1 DOCKET NO. 50-285 REVISION 0 2/22/80 I . l l I I I
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I TABLE OF CONTENTS I PAGE 1.0 Introduction. . . . . . . . . . . ..................1-1 2.0 Safety Features and Design . . . ....... ......... . . .2-1 Table 2.1 Summary of Safety Features in addition to those Ad. dressed in the Topical Report EDR-1 for Refueling Area Crane . . . . . . . . . . . . . . . . . . . . . . 2-3 I Table 2.2 Summary of Regulatory Positions in Addition to Those Addressed in The Topical Report EDR-1 for The I Unreplaced Structures and Components for Refueling Area Crane . . . . . . . . . . . . . . . . . . . .
. . 2-8 3.0 Attachments I 3.1 Appendix B Supplement to Generic Licensing Topical Report EDR-1 3.2 Appendix C Supplement to Generic Licensing Topical Report EDR-1 I 3.3 Omaha Public Power Bridge Crane Seismic
- Qualifications
- I I i I - - - - -
E - 4 g 1.0 Introduction The Refueling Area Crane located in the auxiliary building and used for lifting and transfering the spent fuel cask utilizes a single-failure proof handling arrangement. I I l l I l 1 I I i I l I 4 I l 1-1 lI. _ - . __ ---
I 2.0 Safety Features and Design The special safety features incorporated into the design of the main hoisting system of the refueling area crane precludes a cask drop accident by pre-venting a load drop in the event of a single failure in the hoisting or break-ing systems. Detailed information regarding te design of the Fort Calhoun single-failure proof main hoist may be found in the generic topical report for .I nuclear safety-related X-SAM crane EDR-1, Revision 1, submitted by Ederer Incorporated. The hoist design meets the requirements of both the propriety I version, EDR-1(P), and non-propriety version EDR-1(NP), of Ederer's Topical Report. Attachment 3.1, the Appendix B Supplement to the Generic Licensing Topical Report EDR-1 is a summary of plant specific crane data supplied by Ederer Incorporated for Fort Calhoun. Additional pertinent safety and design features on the new single failure proof hoisting system are listed in Table 2.1. Attachment 3.2, the Appendix C Supplement to the Generic Licensing Topical I Report EDR-1 is a summary of regulatory poitions to be addressed by the applicant for Fort Calhoun. Additional pertinent safety and design features on the existing bridge structure are listed in Table 2.2. l I lI lI - I 2-1 , *g 9i<L.
.> , 3 P. o. sox 24703, SEATTLE. WA 98124 j i 2931 FIRST AVE. S., SEATTLE, WA 98134
.i}.k%.&?h) SUBSICIAR Y OF FORMAC206INTERNA 422 4421 TIONAL., INC h1GNERS AND MANUFACTURERS OF CRANES AND HEAVY MACHINER I
l APPENDIX B SUPPLEMENT TO GENERIC LICENSING TOPICAL REPORT . EDR-1 i
SUMMARY
OF PLANT SPECIFIC CRANE DATA ~ g SUPPLIED BY EDERER INCORPORATED FOR I OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION I AUXILIARY BUILDING CRANE MODIFICATION !I . 't i . . E
!I REVISION 1 2/19/80 i
Approved. C. W. Clark,.dr. / Engineering Manager
;l ,.
I ~
Revision 1 2/19/80 1 EDR-1 APPENDlX B SUPPLEMENT SUMM ARY OF PLANT SPECIFIC CRANE DATA SUPPLIED BY EDERER FOR FORT CALHOUN STATION AUXILIARY BUILDING CRANE MODIFICATION TABLE OF CONTENTS AND REVISION STATUS _ Description Pooe No. Revision Title Page i 1 2/19/80 ~ Table of Contents and Revision Status it ' i 2/19/80 Topical Report Section Ill.C(C.I .o) 1 0 2/15/80 ill.C(C.I.b) I til.C(C.2.b) & lit.E.4 I I!!.C(C.3.e) 2 1 2/19/80 til.C(C.3.h) 2 - Ill.C(C.3.1) 3 1 2/19/80 lli.C(C.3.j) 3 III.D.I 4 1 2/19/80 Ill.D.2 5 I 2/19/80 Ill.D.3 - 5 lit.D.5 5 111.D.6 5
, Ill.F.I , 5 6
1 2/19/80 5 - I P I I
g y im M MMM OMANNECYCkDNS
' g g g W W W IED Y EDERER FOR FORT CALHOUN STATION AUXILIARY BUltDING CRANE MODIFICAT .,-
Regulatory Topicol. Position Report . Information to be Provided Section Specific Crone Dnto ' C.I.o Ill.C (C.I.o) 1. The octuoi crone duty classifi- 1. cotton of the crane specified by The crone will have o Class A-1
. , the opplicant. crane duty classification in accor-donce with CMAA Specification #70.
C.I.b lil.C (C.I.b) 1. The minimum operating tem- 1. perature of the crane specified The minimum operating temperature by the opplicant. for the crone is 50 degrees F. C.2.b til.'C (C.2.b) 1. ill.E.4 The maximum extent of load I. motion and the peak kinetic The crone will be designed such that energy of the load following a the maximum load motion following a drive train failure. I drive train failure is less than i foot and the maximum kinetic energy of the load is less than that resulting from I inch of free fall of the maxi-mum critical load.
- 2. Provisions for actuating the 2. Provisions for automatically actuot-Emergency Drum Broke prior to troversing with the load, when ing the Emergency Drum Broke prior
! required to accomodate the load to traversing with the load are not
- motion following a drive train, required since the maximum amount failure. of lood motion and kinetic energy con be occommodated by the facility design.
p
, . . . . - - ~
g g--g m ~ g mM M M M Y Paga 2 of 6 EDR-l APPENDIX B SUPPLEMENT - l
SUMMARY
OF PLANT SPECIFIC CRANE DATA SUPPLIED BY EDERER , ! FOR FORT CALHOUN STATION AUXILIARY BUILDING CRANE MODIFICA ,' Reg 0 tory Topical 9, 9 Report Section information to be Provided Specific Crcne Data ! C.3.e Ill.C (C.3.e) 1.
' The maximum cable loading foi- 1.
lowing a wire rope failure in The maximum cable loading following terms of the acceptance criterlo a wire rope failure will be less than established in Section Ill.C the maximum allowed by the accept-(C.3.e.) once criteria established in Section
' lil.C (C.3.e). In addition, the lead line stress with all ropes intact does not exceed 10% of the published breaking strength of the rope while holsting the maximum critical load.
C.3.h lil.C (C.3.h) 1. lli.E.I l The maximum extent of motion l. The crane will be designed such that and peak kinetic energy of the load following a single wire rope the maximum lood motion following a failure. single wire rope failure is less than one foot and the maximum kinetic energy of the load is less than that resulting from one Inch of free fall of the maximum critical load. i e
EDR-1 APPENDIX C SUPPLEMENT *
SUMMARY
OF REGULATORY POSITIONS TO BE .
' ADDRESSED BY TIIE APPLICANT FOR FOBT CALIIOUN STATION UNIT NO.1 AUXILIAttY BUILDING REFUELING AREA CRANE (Continued)
Regulatory Topical Positon Report Information to be Provided Specific Plant Data Section C.1.d 111.C(C.1.d) 1. The extent welds joints 1. Nondestructive examinations of the in the crane's structures, existing bridge structure were not which are not being re- required by existing regulations at placed, were nondestruc- the time of bridge construction. tively examined, and Ilowever the bridge structure has been derated from 100 tons to 75 tons, the X-SAM system provides additional overload protection, and the inspections of the existing
. structure described in C.1.b(3) above are adequate to ensure the structural integrity of the existing bridge.
- 2. The extent the base material, 2. The weld joint Feometries used in the at joints susceptible to lamellar existing bridge, structure are not tearing, was nondestructively considered to be susceptible to examined . lamellar tearing.
EDR-1 APPENDIX C SUPPLEMENT -
SUMMARY
OF REGULATORY POSITIONS TO BE . ADDRESSED BY THE APPLICANT . FOR FORT CALHOUN STATION UNIT NO.1 AUXILIARY BUILDING REFUELING AREA CRANE (Continued) Regulatory Topical Positon Report Information to be Provided Specific Plant Data Section C.I.e 111.C(C.I.e) 1. The extent the crene's 1. The crane was not used for making
! structures, which are any major construction lifts and has not being replaced, are been de-rated from 100 tons to 75 capable of withstanding tons . All past projected use of the the fatigue effects of crane at a maximum loading of 75 cyclic loading from previous tons is well within the cyclic load-and projected usage, including ing capability of the existing crane any construction usage. structure.
C.I.f 111. C(C.1. f) 1. The extent the crane's .1. The material thickness of the exist-structures, which are ing bridge structure are such that not being replaced, were Sub-article 3.9 of AWS D1.1 does post-weld heat-treated in not require post-weld heat-treatment. accordance with Sub-article 3.9 of AWS D1.1, " Structural Welding Code." , i C.2.b , 111.C(C. 2.b) 1. Provisions for accommodating 1. Administrative procedures will be 111.E.4 the load motion and kinetic used to assure that a minimum of energy following a drive 1 foot of clearance is maintained train failure when the load between the load and surfaces that is being traversed and when can not withstand the kinetic energy it is being raised or lowered. associated with 1 inch of free fall of the load involved. The surfaces, which will support the load, are de-signed to withstand a minimum of 1 inch of free fall of the load involved.
M MM
- EDR-1 APPENDIX C SUPPLEMENT .
I
SUMMARY
OF REGULATORY POSITIONS TO BE ~ l ADDRESSED BY TIIE APPLICANT FOR FORT CALIIOUN STATION UNIT NO.1 - AUXILIARY BUILDING REFUELING AREA CRANE (Continued) Regulatory Topical Positon Report Information to be Provided Specific Plant Data Section , C.2.c 111. C(C. 2.c) '1. Location of safe laydown 1. The cask loading and decontamina-areas for use in the event tion area as well as the cask rail repairs to the crane are re- car, can all be used as laydown quired that connot be made areas for use in the event that with the load suspended, repairs to the crane are required that cannot be made with the load suspended. C.2.d 111. C(C.2.d) 1. Size of replacement com- 1. The replacement trolley components ponents that can be brought will be brought in through the rail-into the building / containment road bay, which means that any trol-for repair of the crane without ley component can be brought in to having to break its integrity. the cuxiliary building if needed for crane repairs.
- 2. Location of area where repair 2. Repaic work, involving heavy lifts by work can be accomplished on non-single failure proof equipment, the crane without affecting the can be safely accomplished on the crane safe shut-down capability of when it is positioned over the railroad the reactor. bay. There are no nuclear safety restrictions on crane repairs that do not involve handling heavy components.
- 3. Any limitations on reactor 3. There are no limitations on reactor operations that would result operations that would result from from crane repairs. crane repairs.
u
&~M}" N N M M M M N M ' 'M ~'M'M M WWMH EDR-1 APPENDIX C SUPPLEMENT .
SUMMARY
OF REGULATORY POSITIONS TO DE '- ADDRESSED BY TIIE APPLICANT - FOR FORT CALIIOUN STATION UNIT NO.1 . AUXILIARY BUILDING REFUELING AREA CRANE (Continued) Regulatory Topical Positon Report Information to.be Provided Specific Plant Data Section C.3.b 111. C(C .3.b) 1. The, design margin and type 1. Each lifting device attached to of lifting devices that are the hook to carry critical loads attached to the hook to carry will support a load 3 times the critical loads, load (static and dynamic) being handled without permanent de-formation. C.3.t 111.C(C.3. t) 1. The extent construction re- 1. The construction requirements for
~
quirements for the crane's the cranes were the same as for structures, which will not be permanent plant service. replaced, are more severe than those for permanent plant service.
- 2. The modifications, and in- 2. No special modifications or in-spections to be accomplished spections are required to convert on the crane following con- the crane from construction use struction use, which was to permanent plant service, since more severe than those for the requirements are the same.
permanent plant service.
EDR-1 APPENDIX C SUPPLEMENT ,
SUMMARY
OF REGULATORY POSITIONS TO BE
- ADDRESSED BY TliE APPLICANT -
FOR FORT CALHOUN STATION UNIT NO.1 - AUXILIARY BUILDING REFUELING AREA CRANE (Continued) Regulatory Topical Positon Report Information to be Provided Specific Plant Data
. Section C.S.a III.C(C.S a) 1. The extent the procurement 1. The procurement documents for the documents for the crane's existing bridge structure did not structures, which will not be invoke 10CFR50 Appendix B since replaced, required the crane the crane structure was built prior manuracturer to provide a to the issuance of this federal quality assurance program regulation. Ilowever, the bridge consistent with the pertinent was built to the Ilarnischfeger quality provisions of Regulatory control and assurance procedure in Guide 1.28. effect at the time of construction.
This document covered items such as procurement control, receiving control, material storage and handling, in-process control, inspection and testing, packaging and shipping, drawing and charge control, tool and gage control, non-conforming material control, and corrective action measures. e I f I l
lI- . ATTACHMENT 3.3 .
. OMAHA PUBLIC POWER BRIDGE CRANE SEISMIC QUALIFICATION ,
ll REPORT OF FINAL ANALYSIS ; 'I , ,I j$I"f*j - E .
- ,oi g /h k i
- g I
- PREPARED BY TRENCHARD ASSOCIATES - FOR . EDERER INCORPORATED g. Prepared by k+
- Date 3/2/ /Bo
.l v 1 g * .. . s . 2.
*.^*.,'
l . I
i i . l . - i . i 1 TABLE OF CONTENTS , i . i . i - l INTRODUCTION . 1- . CRANE DESCRIPTION - 1 DESIGN. CRITERIA !I 1 LOADING CONDITIONS - - 1 2 MATHEMATICAL MODEL - 2 COMPUTER PROGRAM AND ANALYSIS RESULTS 8 CONCLUDING REMARKS , 22 lg FIGURES 23 'E APPENDICES - l .I Calculations of Crane Properties I . I . - I r .
- .w .
__ _- 1 I..- . . . . .. . . _ ,
.. . ..s
.~ .-
E . \ l 1
~~~
E -
. INTRODUCTION ~
, Response spectrum analysis for the Omaha Public Power Bridge Crane has..been carried out, using 'a modified version of the SAPIV-i computer program (TRESAP 2)'; this modification 'has been made by Trenchard Associat$s so that dynamic analysis corresponds to
- current prac,tice and, in particular, N.R.C. Regulatory Guide 92.
The ' modifications made to SAPIV are described belo'w together I . with the dynamic analysis method, the. modeling procedure and
- the analysis resul'ts.
It is shown that the crane is adequa'te ' to resist both the Design , Earthquake (D.E.) and the Marimum Credible Earthquake (M.C.E.) loadings. _ . This analysis assumes that' lateral. slip. of the end trucks ~ does not occur; use of SAPIV is, in fact, more straightforward when slip does not occur. i Stres'ses calculated on this basis are - higher than those 'which would oe "found if slip did actually ' " take place and the procedure 'is thus. conservative. - (a) CRANE DESCRIPTION
~
I ' The crane. is a standard 2 girder bridge. crane,' the l' .
.. . .- . . .- , . . . - ~
edd truck sheels are outside of the girder center lines. - The live load capacity. is 75 Tons.- I (b) I . . DESIGN CRITERIA Allowable stresses are:. NORMAL OPERATING + D.E' . - 75 x 36 e 27 ' k.s .1.
~
3 c e l I - NORMAL OPERATING + M.C.E. - 36 k.s.i.
.. 5 1
l
,q -
g .-- g , l . . (c) LOADING CONDITIONS l - l *
- 9 loading cases have been analyzed for both D.E. and l .
M.C.E.. events. , These cases are for the trolley'at midspan, . - quarter span and extreme position with hoist cable. lengths of-91 inches (high hook), 492 inches and 893 inches (1ow , , , l hook). (d)- MATHEMATICAL MODEL
.I Mathematical analysis has been. carried out using - -
4 . response spectrum analysis for a lumped mass multiple-degree of frdedom system. The basic analysis procedure is
~
first described and then the'particular'model"used for the . crane is described. I (1) Response Spectrum' Analysis of Multiple Degree of Freedom Systems , i The procedure used is the classica'l normai moile .' analysis which is outlined below.. Details of the
'I computational procedures used are. described subsequently.
I .
.The system 1s modelled as'having lumped masses, previous work having shown that the use of consistent j .
mass matrices for'distribhted weights is usnee'ess ry.
~ ~
1 .
. y .l_d .
z .: .'
. =E ,
. i
. ;*- e i
. i The 5)asic equations of ; notion' are thus -
b
- __ S * %
- ngg
,_ ,r. , ;._.
where . . . l._.,- . S 'is th'e mass matrix (diagonal)'
~
l , Q fis the dampingzitatrix .
\h .is the stiffness matrix
,(_ is the matrix of nodal displacemants kisthegroundac'celeration(scalar) *
{ and n is the column matrix of direction cosines - between the direction of the ground . motion and - the nodal displacement. W ' The first analyt'ical step is the. computation of - mode shapes and frequencies. Since damping is not I involved in this procedure and since, in addition, t'he response spectrum takes damping into account, damping . will not be explicitly, included subsequently. - I In assembling the stiffness' matrix, it is convenient to use the direct stiffness method and there are no ' . masses associated with rotational displacements.. .For this ieason it is convenient to reduce the stiffness matrix to eliminate rotational displacements from the- \I i . equation of motion. i l ll . Thus iet . ,
.f ). -
_. ht ' I
- . . .. . 0, . . .
lI
. .- ..a :4 : ~ - .
i . - where the zero masses are arranged after the nonzero masses in the diagonal mass matrix, and let the homo-geneous equation of motion be ,
'I : .*
., =-7 .
; , , , e
.7 ..
. b S..,, t + k_il .. $i2.
5
%'I
=0
- i. .- 1~2< I. k- s, i l<
4 . ,
,, 2,
{ %s \g
- 3
,I where ( are the translational nodal displacements i and Qt are the rotational nodal displacementis. . Then
-t T lI St
= ~ k t t.
Ns CL , - i
~
and the equat' ion of motion foi+ , becomes
- Mi k i cti =
-t
~ . .M__i n et3 where , . -
I
~
Mi
=
k:.'u
- k i% kat
- _iL Mode, shapes ,and frequencies are now determined -
from'the eigenvalue equation I g- ,
- ca , ' t3i + i.< i m o .
and the normal modes are given by . I
~
~~ ~
i
- g .
.~ -
l
=.*
t I*
'. T.
~
i . - where. - are 'the eigenvectors which'.are deter $ nines !
'~
l together with the freque'ncies Q t '
?l The equation of motion is now written in normal ' -.
is fom .- J * ~ l . j t s ~
+ ..
cl . - u 's ) 1 -
~4rN n gt ~
l The above form for the normal equat' ions of motion . j u is convenient since h .are normalized so that ' 4r y ,. 4 ,. 3 - . . - ,li where. I is'the unit matrix. tI l , Solution of the equation of motion is. nows' traight j forward using the response spectrum method.since max'imum[ displacements for a single degree of fre'edom system lg
!3 due to a given ground motion ..
CQ may be obtained directl" from the spectra, i.e. , the maximum dis-i
- ~
pliteement, from'the equation ( -t..Awt + Et* Rs
- l
\ ~.
(with homogeneous boundary conditions) is .
.-=
g
,- ' . ; - - :. s.4 '~ ~. ~ *. . ^
- l .
E . , , i . where SahW) is the spectral displacement for. given damping ratio,.
.,' h , and' frequency, M . , l ll Al==>
~
SA 3
,4 . -
k is the response spectrum ac'celeration.
~
where Thus, for the equation. , 4 . . l . i
$+A
% h+L g=- M,_ . O g g ,
!l i . - the _ maximum response for the b mode is '
,I
\ - ic '
.sx (x.i w) 9.; t = c where I.
I i l pc= @M
~
~~u n . .
1 . Nodal displacements are 'then computed from - . - ca -
~
1 *
' 14.
I .
~
~ .
l for each mode and . Cs1 .~n ~t- CD E S- s.. .= - %2 . kn (, - g . E .
-v- - - . . - . - _ - _
..... ,c 2.- .- ,
Nodal accelerations are similarly given b'y l 8 -
$O. -
...a
, ,. Q -
-..~>
I i
' Combi' nation of modal effects is by the square root , . ,
t
- of the sum o[ squares for each mode except for modes with frequencies less than;.10% apart for which'. algebraic addition is used. .
Member stresses are similarly obtained separately , for each' mode, from member _ displacements and total
~
stresses 'are obtained by combining modal effects exactly , as foi displacements. l - l i Vertical,. North-South and East-West ground mot; ion ' i are considered separately'. The calculations differ i only by their response spectra and direction cosine I. matrices. Displacements and stresses due to t;he three ' I ground motions are combined by taking the square'ro'ot of the sum of squares. (2) Structural Model '
' The model used is shown in Figures 1, 2.and 3 for ,-
.each trolley position considered. These Figures have
~
-3 .
been generated.by a plottirig routine used'with TRESAP 2 (see below). Members are 3-dimensional beam elements l
iE ;- _ =. ... :. _. . I 15 .
- a. .
l .- -
- except for the. hoist cable which is an axial force i
j - member. Sect' ion. property calculations'are given in -
~
~~~ -
Append.x 1. . f.~ 'The hoist cable stiffness and hook weights are -
.f
. _.= ..-- .
~
omitted in 'the horizontal' directions becaus.e of the low
~ '
i l
-frequency of pendulum motion of the cable. TRESAP.2 -
I computes' equivalent lumped weights due to member- - - . weights on the' basis of ,inp'ut cross-se~ctional areas. . l To include the, weights of stiffeners and, for girder.A,.. .' l - the walkway, input densities have been appropriately modified.- .
~
!g (e) COMPUTER PROGRAM .AND ANALYSIS TECIDTIQUES lg - - 1 . . The response spectrum analysis was performed with
- I TRESAP 2, whleh sets up the eigenvalue equation ,
I
~ ~
DS+ E%
~
- - (1) where $ . is 'the. diagonal mass matrix and $ .' is.the I structure stiffness matrix, assembled from the element I -
stiffness matrices.
.at a node by summing The program computes the lumped masses the mass of all members connected at that node to any concentrated mass assigned by the user to that node. With the mass density, the cross sectional area and the length of each member defined, the mass of each j member is automatica11'y computed by the program. Rotational i
inertias at the nodes are assigned values of zero'. The
~
< k 2' ..- J^ ?-i
>E user, however, may ' override' this and input nonzero rota-I
.tional inertias at any node. .
,The eigenvalues and eigenvectors of eqn. (1)-are then
' 'coniputed by ,TR SAP 2 which uses either the 'det'erminanti search .. ,
., . q._
,I method or the subspace iteration method; the method se'lected - .
l i. dependsi primavily on the bandwidth and size of the structure , , stiffness matrix'. ',$ For a discussion *of these methods, one is referred to ' Numerical. Methods.in. Finite Element- , Analysis' by Bathe and Wilson.
. .l J
Since the' behavior of this systiem is governed largely by the lower. modes, one need not compute all the frequencies and mode shapes; for some '.large' systems, the solution of tihe intiie eigenvalue~~ problem may be economically prohibitive. l The number of frequencies and mode shapes computed by the program is determined by the cut-off freiuency l specified . .- by the user. In this analysis , a cut,-off frequency of- , 33 hertz was used. SAPIV then computes the maximum modal responses - or displacements as follows: I Maximum modal displacement, f% c.x , due to the f x-direction excitation for the T mode of the structure is:
- g. Tw L .= %3Lk)% W s M .
I . l
E
' . J-:R5.g_. , ;.LW:" d' ' T.~~ -
f*" ~~ ' * * * - E ,
. .~ *
-11 . , ,
I ~ displacements, nodal accelerations and element stresses .
~ '
. can then be computeil. Since modifications were made to
.those SAPIY sections computing these responses, one is I. ,
. referred to .' Program Modifications ' below. -
(1) Program Modifications
^ . - -
I , In this analysis, a version of SAPIV as modified by Trenchard Associates (TRESAP 2) was used. Following.
's i a presentation of the differences between the -
~' '
original a~nd modified versions of SAPI7. For spAcific programming details, one is referred to the copies of I
- the . original and modified subroutines 'in Appendix,II.* ,
.For a response spectrum analysis, the original -
version of SAPIV accepted only one response spectrum. . . . with the user specifying the fraction of the loading - , to be assigned to the global x, y and z directions; the response spectra in each of these directions thus had to be proportional to each otihes. . The progra51 was modified so that three distinct response spectra corresponding to the three orthogonal directions could be handled. .See subroutines 3PECTR AND SDMAI. I . Also, modir_ications were made so that the user - - I . could input acceleration or displacement versus log frequency spectra; the original program only accepted . displacement or acceleration'versus period spectra.
~ ~ *
- Appendix II is not included in this report.
I _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . 1
~ .h.-:;-; , ... . 2.a_:. . L- 7 __ .
~
g . g =. Modifications were also made so that the user could input. l several' sets of spectra in a single run. With the original program, to perform an analysis with two sets of response spectra, I for example, requir,ed three computer runs. In the first job the frequencies, mode shapes, element stiffness matrices, etc. would be computed and stored on tape. Then for each set of response spec-tra, the infomation stored on tape in the first run would be retrieved and a response spectrum analysis performed. In the modi-fied version of the program, this restart capability has been retained, however the user may now input any-number of sets of response spectra. These sets of response spectra can be of different types; for example, one set could consist of accelera . l tion versus log frequency spectra while another set could consist I of displacement versus period spectra. One is referred to sub-routine RESPEC*for these modifications. Extensive changes were made to those program sections com-I puting and prin. ting the nodal displacements. The original version of SAPIV printed only the displacements corresponding to each mode shape. In the modified version, the x, y and z direc-tion excitations are considered to be applied separately to the structure; a R. M. S. displacement vector is also printed.
- Subroutine RESPEC is not included in this report.
I - I
g _. .+ ~ c _ w -
..,. v g .
~
~
A brier outline of the procedure is as follows. E' ~ The. x-direction excitation is first applied to the -
. ~
E ., structure; the amplitude for each.mode
~. .
shape is 'then., .
~
computed.
' The amplitude times the corresponding 'eigen-E ..
vector yields the displacement ' vector for that m. ode. , The displacement vectors for ea'ch mode. shape' are, then' combined.by the square root of the sum of the squared ' (SRSS) method, excepting.those. displacement vectors I
. corresporiding to freq'uencies, within 10% of each other; l
the absolute values of these vectors are first summed. ;- together and then combined'with the others.using the ,
'SRSS' method. '
~
l
. t 1 Da* 82 X
~
l1 D m %gu 8-
* '9 k h W, moods,/
o W 3 nn
% Tnow /c */.m op hCu mh cq
~2 ,
g + N- N==ci e - - ' vb.,, ~.Jc ~ 2 i Responses, corresponding to such closely spaced l l frequencies, are essentially'in phase and as such .
~
should be summed. - This procedure wa's incorporated into the' program I - by essentially making use of an array 'ISUM' (see the modified version) fil' led with 'FLAds' indicating which frequencies are within 10% of each other. 1
I .
- .....; q.c...
8'- r -
~
The result'ing displacement vector is then printed by calling subroutine PRINTD; subroutine PRINTD was modified so as to read this vector off its new location .
" ... ~
,on tape'and,new =--
format
.- statements were adde,d to improve the -appearance of the ostput. This procedure I-is then .
- ~,' . ~; :? .-
repeated.'for the l. oadings in the y.and z directions. ll Finally .the :. =- . displacement vectors correspon' ding to x, 3
r -
~
. y and z . direction excitations are ' summed byl the SRSS .'
method t'o yield a ' total' displacement vector.' i
. 1.
~
.1 hrs, Torw *
- 2 a -' ,
- % D *"8 : A ine ,y +
s S.D*^ . ' '. L (See subroutines PRINTD and SPECTR) SAPIV was further modified to compute and ' output ' . .the.no'dal accelerations. The procedure is exact:1y 'the i same as' that described above except that the maximum - nodal accelerations, c[,, c,x % i rather..than the maximum nodal displacements, {,gd% are used to" scale the eigenvectors. A number of modifications was made to subroutine STRESR. ' ' ' ' The original version of SAPIV computed the element forces as follows. 1) From the computed frequencies, the modal participation factors and the response spectra, the maximum-respo'nse or amplitude of each mode shape was determined. The displacement vector for,each mode l .
_ . - - . q % g ,.= .
... = .:.. .. ..
- =
..-~:--
.=..-.=- ~
r- _y-
- l .v -
was then' comput.ed by multiplying the corresponding
~
eigenvector, by its maximum displacement or amplitude.
' These vectors were then stored on tape.
~
2h Ea'cli' element i - . matrix,.'b4.. was' previously multiplied by its trans-
' formation" matrix " .h ' 3.n subroui:ine NEWBMi the re'sulting matrix, (d.M s.E ev wais sk;ored on tape. The elem5nt
' l
- forces' ~ were then obtained by retrieving these matrices, Idg7 from tape and multiplying them by the app'ropriate global displacements.
, This was dbne for inach mode.-
- 3) For each element, the sets of element forces corre l '
sponding.to each mode were.then combined by the;SR'SS, method. For example, there are 12 forces associated j I. with a 3-dimensional beam element; at each end.there . are an axial force, P ., two shears M and %; , a torsion T and.two moments M t and Ms. For each mode there is a set of these forces. SAPIV will output a 1 l . . single s'et of 12 forces for each element computed as follows. - . . - N O D E.S . l r = )q. T.
~-.
W 2 i M g V, w 3 4 g - w = i- - I l 4 MOD P.!
= Vt m Vg
= N wg d .
1 : l . g e.6.
l --:J.~.gC:s,,.f:L; .X - ' a
.=
~
,.i. .
, 4)
The maximum stres,1 at each end of the beam element I. ~ was then c'omputed by hand as follows. 4 .~.
' P % . .Ms -
PAKw ~- A . ' 11
. .. .7,- :.
k 5.# 2
.v.- . - . . , . .
I. T. # '
~ TRESAP 2 comp'utes the stresses as follows. . .
la) The max,imum response or, amplitude of each mode shape is com_puted separately for the .x, y and z direction excitations. These amplitiudes ai'e stored ~ in a matrix *as 'follows; l - inode x ,- * . I* 1 l (s 3 I - '
~
I amplitude for eigenvector ~
,' l corresponding to mode 1 with-only the x-direction excitatic applied M t The displacement vectors corresponding to each mode are not directly computed. However, the dis-I placement vector for any>. mode under either x,y or z . -
direction loading can be quickly determined by lookin g up the prope'r amp 1'itude,in the above matrix and scaling the appropriate eigenvector. '
.- ~~
,2a)
Againasinstep2'above,.theelement~Ee.r . i matrices are retrieved off tape. Subroutine TEAM I was modified so as to read'the'section moduli.of each. element off the data cards and 'then to store *
'l I ----
.I l
- the cross sectional area and section moduli or
- each. element along with its $e.Tmatrix.
. These
- ' ' matrices are then multiplied by the proper global -
. displacements to obtain the . element forces f' or each
~
mode and fo'r the x, y and z direction excitations 3a) l . The stresses for each element are 'tinen
~
summed as follows.:- ' . . !I -
- E
. .Ho%ts I
5 P I Mg O "^* - * ' X * ~ [ Mg ! ) 1"! u E'+'3*'> - a' 4-1-sh4 m.
~
Analogous to the. procedure used in the computa - tion of the nodal displacements, those stresses corresponding to, frequencies within 10% of each other are first summed together and then combined with the others using the
'SRSS' method.
U and O' m s, . 3 , are similarly computed. Then ' ' I , kAs, *
't-1 L
, max-x
- O max- *U .Mu - a. ,
5 - New format statements to ifnprove the appearance of the o'utput were written. These statements e eliminated the need for subroutine ELOUTR. Lastly, SAPIV'was modified to create and I e
) . catalog a data file, on command, for the; program .
'PERSPO'., described below. * '
~
Drawing program 'PERSPd' ..
~ '
Program 'PERSPO', developed by Tredchard Associaties, is primarily ~an interactive pregiam
. . capable of generating 3-dimensional perspective . - .
drawings on either a CALCOMP, or GOULD ELECTROSTATIC '- PLOTTER. - Commands to rotate, translate ~ or . scale the \ . - structure are incorporated into the program. . The. g user'also has control over: 1) The numbering of the nodes and members, 2) Partial" drawing o'r the ~ structure, 3) Selection .of different pens for drawing and 4) Selec. tion of differen't vani:ag's points in the generation of perspectives. The program is particularly useful in detecting errors in the member and n6de' number'ing. -
~ '
(f)~ ANALYSIS RESULTS ' I The static analysis of SAPIT was not altered and, for . I . . this case, element axial forces, shears and moments.'are . i I given from which stresses must be computed by hand.
. As described above, stresses for dynamic analysis are given-directly. . . . . .
l l l
~
'l 'g ess (1) Interpretation of Computer Output As stated above, the beam geometry for each of the 3 trolley configurations given as shown in Figures 1 through 3. Figures 4 through 6 show the~cooresponding node and element numbering.
- More detailed information of element numbering is shown in Figures 7 through 9 where each of the three beam groups is separately shown.
I FinaHy, in Figure 10, the spatial coordinate system and boundary elements have been shown. Boundary elements are used so that i wheel reactions were obtained directly from the output. The boundary elements are chosen to be very stiff so that they do not affect frequency computations. Verification of this is seen by noting the displacements of nodes 1, 4, 20 and 23.- All weight and section property computations are given in - Appendix I and are not discussed further here. It is to be noted that there is no significanc'e to the hoist cable stress since the cable is loaded as an axial member whose stiffness is chosen to agree with the hoist cable stiffness. The cable strand force is I obtained by dividing the hoist cable axial force by the number of strands (16). The D.E. and M.C.E. values are given in Figures 11 and 12. I '
~
I - i I~ . E
(2) Summary of Rrsults' 4 '
- D.E. -
Value L a. I ,
.c l Marimum girder stress (k.'s.1.) '
17 71 91 Maximum ver.tical
. reaction .per girder (K) .
. .:..' 27. 2
!E . .
165 5 91 6.42 :
\ E. Maximum late'ral reactiori per girdei -(K) 24.1 \
. ~492 13.6.
Maximum longitudinal react' ions per girder (K) 40 5 '
' 91 -27.2 Dynamic boist.. cable force (per strand) ('K)' 3.8- '
- 9b.
27.2! i ' Girder displacement at trolley (in.) x-direction
. . .i 30 91 !l
- 27. 2. :
Girder displacement at trolley (in.) z-direction t .15 91 , 27.2!j I
- Maximum hook displacement (in.) .
~
.42 -
893 27.2!I \ M.C.E. - 7 Value L a Maximum girder stress (k.s.i.) i 20.00 '. 91 27.25
~ Maximum vertical reaction per girder (K) 171.6 , 91 6.42 Maximum lateral reaction per girder (K) 29 3 492 13.63 Maximum longitudinal reactions per girder (K) 55.0 91 27.25 Dynamic hoist cable force (per strand) (K) 5.1 -
91 27.25 Girder displacement at trolley (in.) x-direction '
.42 91 Girder displacement at trolley (in.)
27.25 z-direction
.20 91 i
27.25i) Maximum book displacement (in.)
.60 893 h
27.25 , i
- \
't
!E
- 5
)l (r) Concluding Remarks Girder stresses are substantiaHy lower than allowable values and
- the structure is clearly safe for both D.E. and M.C.E. conditions.
? j .. - {-
~
l - !I 1 !I 4 lI 4
- I 1
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- l E - R. J. Evans P. E. , Ph.D.
'g THE TRENCHARD ASSOCIATES !I
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r v42. , a I l
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. :. . ::. = . =. . ..
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. - Figure 1 ' Crane GeometT7 -
Load Central l l
E. - Ee O I e e. g . . E * -
. 8 I
g . . g W E . I l
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Pigure 2 - Crane Geometry - Load at Quarter Span . E *
- g .
1 i I. . -24, t .
.
- I
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l - - I i - iI 4 i i l '- l - i I, - I . I ~ I . . t , Figure 3 - Crane Geometry - Extreme Load Position e'
+*
9 . O E .
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= y o ,
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=
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..,, Figure 4 -
Nodal and Element Numbering . Load' Central . l I h ow l l
- tEb%
I E* ~ I .. ;. ..
- - - .r n . . .* .:*; . r. ~ * ~ ~ -
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1 ,. Figure 5 - Nodal and Element Numbering - Load at - J Quarter Span
= )
i . 4 , 5
I.* l l - O 9 I , e * , IO e 1- . . e , '.' E . e
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Nodal and Element Numbering Extreme .
~
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. * +
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s - ~ O
-ee.-
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- s. .
m . l
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1 - - 1 p . I i l M I cw I I . . E. I- - Figure [ - I. Beam Group 1 - O I . g *
- I -.w, e--------- -.-, , , - - - - ,-a ,u , w,,,,---.n--,-.
- - , - -------r , - - - - - - - - - - - - - - , - - - -
e
E. a I. g- . E .- . l - . I I 1 -
; ...==.- .
'I
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- y ,=,
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-E .
I . . . E . I.e
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1 I s- . E e I . I . 1 I _ _ Figure 8 3ean Groun '2'
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?4 E .
B s; I I.. I E ..
\
I I
- ~ -
s, ,I-l : .- . E. -> - I i t .
~
>E . .ln . . . Figttre 9 ' Beam Group 3 Ig . l i l I . E . l
-- =. M: :..f .,.-, :_r*- i
~~~.~*
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I l . t i - l g . y O
- a.
- g-w4
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=
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( g3 l L/ 4y . . 3 l l . 1
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~
F1ure 10 - Spatial Coordinates and Boundary Elements r - .
~
I .
- l. -
l TABLE II ' LINEAR AMPLICATION OF RESPONSE SPECTRAL FOR .08 HORIZ. ACCELERATION EQUENCY ACTUAL ACCELERATION AT 989' 2% OAMPING AMPLIFIED ACCELERATION AT 1064'-3h 5% DAMPING 2% DAMPING ._' 5% OAMPING ICPS) . = . 0.02 .Ecso 'O.020 :. J.~ ~- 0.016 0.051 J-o*l - 0.041 04 2so ._. ~ ' . 0.080
. ?_U - ~ 0.065 0.204 ' cot .0.166
' j. .
,06 It 7 0.180 0.148 ~0.459 oo3
'O.377 *
.D8 11 5 O.320 -
0.260 0.816 ccfs 0.633 10 fo o 0.500 0.410 1.2 75 -@ 1.046
.20 so 1.800 1.500- 4.590 o'58 . 3.825 .40 25. 4.000 . 3.100 10.20 cMh, '
7.905 0.60 147 6.000 - 4.400 15.30 -lol i 11.22 80 g.2s 7.850 5.550 20.02 13 1 14.15 00 g.o 9.250 6.600 23.59 lT7 16.83 2.00 .o.S 15.000' 11.000 38.25 155 28.05 00 c>2T, 19.000 13.000 48.45 h'k. 33.15 00 e.t 7 -
- 19.000 . 12.000 . 48.45 n3; -
30.60- ! 00 en 15.700
~
10.300 .40.04 2G'l1 . 26.27 l 0 c.to 13.000 9.750 33.15 41.t 8
, 24.86 0 e en 8.000 8.000 20.40 .tg '
20.40 0 7 8.000 8.000 20.40 20.40 0 p' 0 0 - 0 - 0' 0 O O O O O OCl 0 0 204
- l% O -
! PC**D M ) ' MULTIPLICATI0H FACTOR
- 2. A M'"" WG
.204 = 2.55 . ,
. Titr- .
l ' l - l Figure 11 - Design Earthquake Horizontal Resperise Spectrum . g . , Note Scale. accelerations by 2/3 for Vertical D.E. 5 ' g l . j . .
. TABLE I -
l LINEAR AMPLIFICATION OF RESPONSE SPECTRA FOR .17G HORZ. ACCELERATION
~
SQUENCY ACTUAL ACCELERATION AT 989' AMPLIFIED ACCELERATION 1064'-3h" (CPS) 2% DAMPING 5% DAMPING 2% DAMPING 57,OAMPING
.02 So.is ' O.'044 0.035 0.076I -Cof - -
0.'131 -
.04 2.5.ol 0.170 0.140 0.292 oo2- 0.240 -
0.687 'cos
.06 N7 ^ 0.400
- 0.320 -0.550 .
.08 to s 0.700 _. 0.550 - 1.202 -cos 0.945..
.10 1o o- '1.100 + +-- .
0.900 1.889 -oG - 1.546 ~ . 6.527 oW
.20 5' o - "3.800 '--
3.200 - 5.496' . ~ .
.40 1- 5. 8.500 6.100 14.600 W 10.478
.60 l 6,1 12.500 9.800 - 21.471 14~6l . 16.833 .
.80 f. "L5 16.500 . 12.000 28.341.Ib3lg- 20.612 35.212I .7.E -
- 1. 0 - f.o 20.500 -
14.000 24.047
.0 oS
~
i 33.000 23.000 56.682 alo 39.506
.0 e.t.C 38.000 25.000 65.271 425i 42.941 . .
.0 o .l-) 33.000 23.000 56.6E2 3 1s I- 39.506
.0 lo-l2. 27.000 19.000 46.376 ! , 32.635
, .0 I cs to 24.000 18.000 . 41.224 So9
'275 j , ' .
30.918,
.0 ' o os 17.000 17.000 29.200 . (% . 2.9.200 '
.0 i 17.000 17.000 29.200 2 29.200 :
) .0 .l 17.000 -17.Q00 29.200 . 29.200 .
.0 -
16.500 28.341 28.341 . 1 0 ci l16.500 16.000 16.000 _ 27.482 *l % 27.482 ll0.0 / .
\
PEh (scc) 2% hANy N C
~
MULTIPLICAIION FACTOR . ,.
.292g = 1.718
.11 g ll 1l Figure 12 -
~
Maximum Credible Earthquake Horizontal Response-Spectrum Note Scale , accelerations by.2/3 for vertical M.C.E. -- - jj
; E. ;... . -
I~ - ll . i .
! 7.- S
l
! - S l
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.- t.~ *:. * -
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l
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l i . I i i e , E . . . s
- l t
d APPENDII'I
-I .
4 i
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I
- I,.
I
- I l
I e. k E .
.I.
I
~
E -
I. . _ ,
~
. . . ~~
~
SECTION AfD MATERIAL PROPERTIES ' . I The section and material properties used in the analysis are below. These properties are reasonably representative of those in the .E- actual structure.-. '- .
. . . . . . + .-
j.
~~ ~ , , * . .
. END TIES ~
E .r ^ l .
,. ; +
l*
.3 , ,, --
4 ; tQ4 ily, , 3
<, - '4,,
_te"4 ec r, ___ EO , e a I , - I
~
E - - I I I 4~ . . i l TO *, ir (I , I ~ I .
i
.+.. T. .: . . .
= .: ;;. -
- L..2
.:_ m.. .
-::::1 :.;:= . .
. =.m ~:.z:L :==~ . --Y-q_= ==i. :. - ?- .
. . m. =..
.~ - --.. .. --i m-- :. 7:L y.
- :6:; .'--KQ V~~==.: ~;~5 fit. V
.L I ..-._.. -. ..Computation .: ' 155~t-
.: . .25.1: . .:.i O :. ri ' .
~~; of shear area 3 2 - ~-
=.-.. . :. .i '. '. .
g.
=.
I
'A J.
._ .sh = 2x18x1 + 30x.5 .. :._ = 51 in .. m. -- .
=. -
.:. . 2
- '~.
~ ..Asv ,= 2x15x.37_5.+,2x30x.5 + 4x.375.==w44.75..ain x :- - :... . 2 . : .-
.~.:: - : . :- :- - .. : .-.;. -
=
..:- S;--;_.
. - . -. -- : ... .=.. =
. .;g" .-- . . . . -- -:.:. W . -
- - * ~~ : ,
7 ;-i.:. .. . : = :.--
~n - : 4_ ::n.: ;:--q. .i._,;.:....5-_ : .:,.-
.c
~
. Computation of. area ; m- .__.-g.-.- -- :-:~.:: .,.:.._ . .;=,jaj a.f. . - -- . .;; . ; .- . :q .
- ===. :- =.::.: .. .
. = . . - :=; n -
... - _,, : n. -
... .. . .. , - .2...=
. . . . . . .7 :
.g-.
.i -
' 1A ..::.
= 18x2 + 2x32x.5 +:2x15i.375 + 30x.5 + 4x.375 = 95.75 in " ^ .
.--..:.- -- :~- . .
- 'q .
~
5 Computation
.;. .: ... . of torsiona'l' ~__/. ; .........l..' -
= :moment...
2 2 " [=2x11.5x32/(11.5+64)=3590in4
- I .
' ~
J
- cor.pu' t ation of location of centroid. IG, and Zy .
I Y = ( 18x.5 + 2x32x17x.5 +.18x1x33.5.+ 2x15x.375x51 + 4x.2375x56.5 - .
'I -
+ 3'Ox.5x58.5-) / 95.75
= 28.1 in .
3' 2 2 ,l I y = 18x27.62 + 18x5.42 + 2x.5x32 /12 + 32x11.1 ] 2x15x.375x22.9 3
+ 2x.375x15 /12 + 15x30.652 + 4x.375x28.42
= 42300 in 4
- 3 1
Zv= 42300/28.1 = 1505 in . i E Computation of location of centroid, I ,h and Z h. I
~
X = 18x2x9 + 32x.5x3 + 32x.5x15 + 30x.5x15 + 15x.375x3.0625 + jlI 15x.375x14.1875) / 95.75 ] = 9.9 in 3 2 i Ih = .5x30 /12 + 2x.5x15x5.072 + 4x.375x18.63
- 15x.375x6.872+ ,
3 I . 15x.375x4.252 + 16x6.932 + 16x5.07 E + 2x18 /12.+ 2i18x1x.932
= 4580 in 4
~
1I 3 - Zh = 4580/9.93 *. = 461. in
] *:
t I Mem e i I -- e
---~.-r-- -~
~.; _+ - .- a , +;
. : :_ =.;3=p::g- . ;;= -: =q~ ..a.:3:
. . H+-:. - -
- . _ - - = . . . = . - .. .. .-. ,e . - .T, .:;. . - .-
- x:.3. ..c=-
.= .m -: .-,.
- W .*-
~ ~'E. -- *'
- L.
END TRUCKS Y. W.J. .-:-9E?-- .?# .
~
;- ,,h .~. { ~ ,.....;*
s ,,. *.: '
-~ '_.
.- gg*
_. y .
- . :-- .,b a
p -
-2: .
l f
- . , C 5.. -
E s.f, -
;;f.
- i. .
. ,2 . ,
~*
?&.} .. -
- .Y;*d~. :: - *
~T- - 2P'* - ' -
%.' . . " . - ; L ~.-
-*.* '[. .
t.y- . ..-:.
- . ::::.:=
.. - ~__.-,. .: .. : .,.- . . .. ----... :: : ..= .:~:-- - . -- --- . - . . . . .-_r. ..-
. : .r~.
. . ... .~:: :. - .
.-- : .. . . .:r *
- .:-~. : = :. - .: . - :. e . .
.. =
. : .; .;._ g. . . '+. x-- .-y_
- ~5 34 , . , , ,- ,
... =:^:5*:4. ..,;-..: ~. .
2 - :- .
- . . 8,...- .
. IS -E, 3 .
.t., ,
E t I -- g, , ,, u ,, .- - - W I aar - 4 3-I 2.as' 3 Computation of shear area - I 2 - A sh = 10x1 + 2x18x1 = 46 in -
= 2x32x.75 = 48 in 2 -
- A sv Computation of area
- A = 2x18x1 + 2x32x.75 = 94 in2 l . .
Computation of torsional moment I 2 2 J = 2x10 x32 / ( 10 + 42.67 ) = 3890 in 4 I Computation of location of centroid, Iy, and Zy . l ' Y = ( 18x.5 +18x33.5 + 2x32x.75x17 + 10x20.125 ) / 94 = 17.3 in I,= 18x16.22 + 18x16.82 + 10x2.83 2 + 1.5x323 /12 + 1.5x32x.32 , ..
= 14000ini g Z y
=
14000/16.7 = 838 in 3 .
?e E ei
.- ; - .;;g.; . ..: :.
. . ;c. :. ._ , .... -_: =. :.-. a .; := ; ..c . : - ..:
._~ . .
. .:. . :.u .;
.....a.. . - .
, .- . ; .=. .. .. . . .
8 Computation of location of centroid, I ' and Z * '
- h. h n.. a. . .
X = 9 in m r.. :.- - - .- .
.2 .. . : "
"" 3 3 ' ~
, _ , _ . . Ih I. 2x1x18
/E+ 1x10 /12 + 32x.75x5.3752 = 2440 in4 "'-;. ~- -
\ ; .. 3. - . . . . . .
. =*: .J
. h . .'- Z' _ 2440 3 - __
-' - 7~
~ E1; h =N.m. 2 + :/ 9 .=_271~.in c :2 _.-
--- . r:. . :. =._.;
L :- -. .
- ..=----
.T_x . . ~ .: :
~ .
., a J :,..j._k ~ .. ' .. ~
~ - '
CABLE STIFFNESS 2 K=ND x 100 - Where N = f of strands
'D =' diameter of strand -
F = rope factor' - Le= len~gth of cable - I . . I Therefore K= L 16 (1.25)2
- x 100 .
e (1.6x10-5) I-- . I ,
, 1.563 x 10 8
'c Take A = 1.0 in2 8
E = 1.563 x 10 p,4 .
~
for SAP IV RUN , I .
~
I
.- , _ , ,. - ' ., - *., 'M, . -
~4 '. MAIN GIRDERS l _
) ? J\ .
i ,..- . l
,. I *
. . J , .~. . J;;,.
I.,
..:.-*~* - . a. '..;'* ;7* *
*~.
**I l. . ,*
g'* ,
. .:--**.** * .; ; . ,.L ' . ..
- .-- , .~.;s;- - -
I
- .r;;;_.c - -
- --.. ==. - -
- - y*, .g. .
..9~ ~
s , I
~
..-; --4: _ .-.
~..-- .: n. ,
I
. ,r ,
[p a - m h .24
. a r- .,
4 . . s 22.< %, I Computation of shear area . 2 - - Ash = 2x22x1 =144 in , 2x70x.3125 = 43.8 in 2
- ~ *
- A =
sv . Computation of area . A = 2x22x1 + 2x70x.3125 = 87.8 in2 , _ ,
' ~
Computation of torsional moment I 2 2 J = 2x20 x70 / (20 + 224) = 16070 in4 . .
' Computation of yI , and Z . -
'..1 y . .; . . ..
I Iy 73344 in.4
= ; . . ~-
~
. Z y.= 2037 in3 . =.:;';E*'. - -- --- ?' ir --
m.. -
. ,: .1 :. . .2. .,. ..- =.. .
.E
. .. . .u- . . --
. ..m._.... *.,_
~,..~ .,
- =*
.. . . . . ~f . . - -....
- .3
. . .:. . z .
E _ .q.. % 4 .-- q e M e,
. q::..
g , p l y.: : ;5 ..._ .(. ..
* :- f: ,
. ,,.7,... .
~. ., .
..~ -.,a ;, L. ~
~
I ~, a, . . .
'".~.**."
;-.-.~.'
.. .,y; :. .; .
L - - - -
- r- '
ae
-.g g ;.7. .. . . . . ..u..=-
~
I.
- . . =-- - .
I ~
- . ;? -
Computation of I . and Z
- h h -
=
. z.. . . -
.. . . . ~
.-- IM I' = 3
. ..=. ~ .-.. h 2x1x22 /12 ..u + 2x70x.3125x10.15632 = 6290 in 'J ~
._- .=.- . .
s-
. a. . -
f.. I=~^...:.*:..I..
. .. .~ .
. . h. .=.
6290 3
/ 11 .
=.,5,72..
in
??- H=~ *~
..; ' : .= . - -
...a... . ,
L .L -
.p +.: y . + =- =
- s. . .. . .
'" ', ~jWEIGHTS AND MASS DENSITIES - '
I
'The mass and weight densities of the rigid beams are taken as zero .
I The cask and trolley weights- and masses are then entered into .the. program as concentrated at certain nodes ( see below ). - 43'.5 K ~ 43.5 112.58 112.58
,o o
kQ "5.176-b - E. Cable . I . C 163 K-(
" 421.84 I. Hook weight = 75(2000) + 13000 ~
~
*h WEIGHT' 7. -
1 I. MASSES . E. : 7 s
. . *- r
* * , -} .; . , ;.. ..
.~ ":
T.*:.
- p *. '... '.
...'--y
^
I
~ .1,
. , , ;. ....'. L-g..
, . ~;.T *. . . ~ . :E m-- * ~s[ .'----*
's ,
~~7..'.- . , ~ .;. m_Q* tg*L.'.' 1~e_ff ; _._ .--. :%2a..
--,.. -;-*. C .: ,.;.:.3 , . . - ...,*-.s.---s
; .. .'s. ,;.e; x - -:- ::;- -~
; 's y - . . %; .
_~_==. . . z. z - . ,9. .,L'~k .; . ..ia: , . .w: --. : % A q;.~ c-. p-'.~r: , Yh I
* **Cfi.#g.' . . . .. N., ' 1. *-$*
~
M,. 2 .. '~. , ;
. . ~ rQ , ?.
.f
,. .%W3[:'m*"I~
;47:. - ~ . -N-
w -~ I' . . 4._ .se..
.. . . - Q-. .1%. , Ya:a 4' apt-,;n :~- : s'1**
. . - m-'.
_, *e _.* M. ~_ - ..**.. . . ..
I Oneofthetwomainskrdershasa12 Ewalkway attached to it. .An . extra web is used for stiffeners. The mass and weight densities of the main girders were adjusted to . . I. incorporate the mass of the stiffeners as well as tfuit of the walkway.
~
- GIRDER "A" l
Total girder "A" weight = 87.8x654x.284 + 12000 + 70x(5/16)x654x.284
, . . 5~i - . = 32.370 lbs. ,
3 Equivalent. weight density = 32370 / 87.8x654 = .564 lb/in equivalent mass density = .564 / 386.4 = .00146 I
~
- GIRDER "B"'
Total girder "B" weight = 87.8x654x.254 + 70x(5/16)x654x.284 = 20,370 lbs. Equivalent weight density = 20370 / 87.8x654 = .355 lbs/in 3 I equivalent mass density
= .355/386.4'= .000918 I
~
TROLI.EY PROPERTIES 3 j,68, 3-sodei,ed as y@ ,
~ Y D l geY /
W / g h g kQ I w
- Cable -
Cable
= .
.* -~--n
_ _ . _ 1. _
1-m'.. '9'
~, *M
^
_- *j
-T- L . ;- _
'._2u .FC. . ' - , .;' . D =L' W: 7. M -' W . ^^~..
, . -Ji t -
i , If '. ~. ;.tr h=~ Etll'h' 5"-.-^f= .* f ' ,b _
~-.
? *~-
. . . . . . . -~
.=.. ._
~ Beams 1 , 2 , 3 , and 4 are rigid ; very large I -
- I 's h v IE and A's are used in the computer run. - l
. g. - l The equivalent spring: stiffness K. EQ is computed as follow;l
. .. _.y. -':_.. _ , ;_--. _ :
^
- . . - .+. ..+f, - . . ' . 6 .i=
- . - .1ik - .
e..=. : .. .._ __ ._;.._1._.. _ . , .. .. . . -. -e. .-<.m... _
~ .:
. . ...-e _. . - .
.s .:.:: . .s....,.
1 . . -
.. ge.1 .. .
- .a.. - - .
g..e . .: .
. 1. -. . "fr"2 1 .
. o .-
.m .... .. , . _ . _ .
a, e y . .. b l
- 6 l Kg = 48 EI / L3 = 48x(30x10 )x4396 / 1323 4
3 6 3 l K2 = 48 El / L2 = 48x(30x10 )x7245 / 168
- ~
' 1 6 K EQ "
= 1.572x10 lb/in - -
3 7
+g a 2K g lI 3
For the equivalent truss el'ement use : EEQ = 1.57h106 WM A = 1.0 in t = 1.0 in . i - 4 I n . J i i i .-- ~- _ , - - - - j E .'- . 3 - s -
.~
1 i .- . . .
; 3 -
.u. .
- J . . .
!,I _ _.= .
. _ - , -.: ~- . , Z . * -j., E , -- ^ ^ , -
- ~
- .e.,
;I a f. W i :5. .-Q L 5 ': W D - ---
i i j !I 1 i !I 4 lI i l 1 APPENDIX D i l l CRITICALITY ANALYSIS DATA l l I I I I I I , I I I I I
APPENDIX D KENO IV CRITICALITY CALCULATIONS FOR OPPD CRUSED CORE RESULTING FROM ACCIDENTAL I HEAVY LOAD DROP OF RV HEAD Program I KENO IV (Ref. A.4) is a multigroup Monte Carlo criticality program written for IBM 360 computers, and is benchmarked against numerous criticality experiments (Ref. A.5). Input data to KENO IV consist of four different sources: card image control data, cross section data, reflector albedo data, and group weighting factors. Principal output consists of k-effective (the aajor objective in this case), lifetime and generation time, energy-dependent leakages and absorptions, energy and region dependent fluxes, and region dependent fission densities, all in report format. Although not utilized in the crushed core calculations, KENO IV I has a restart feature and more detailed report options, and can accurately model any geometry. I KENO IV solves numerically a derived form of the Boltzmann neutron transport integro-differential equation as a function of position, direction, energy, and time. The theoretical derivation and equations are described in Ref. A.4. Model and Calculations I The homogeneous regionwise model of the postulated OPPD crushed core set up for criticality calculations is described in Figure D.l. Hansen-Roac-16-group cross sections were used with the 0.5 energy group default weights I used everywhere. Fuel (i.e., U and U 235 tith other uranium and plutonium isotopes neglected) resonance-corrected cross sections were applied to muimize k(eff). Code description notes that calculated k(eff) is larger when using fuel resonance corrections. Since fuel and poison I configurations are not predictable after a postulated core crush in the vertical direction due to a heavy load drop impacting the core top, a homogeneous mixture of core components was used to approximate the actual core condition. Table D.5 lists the active core atom densities (units: number of atoms per I barn-cm) used in the homogeneous model of the crushed core. The resonance cross section corrections in Table D.6 were used to compensate for the fuel matrix self-shielding. These cross sections were calculated by dividing the potential scattering cross section of the medium by the U 38 and I U'35 number densities. Linear interpolated Sig P were used to weight the KENO IV-referenced applicable cross section. I I I D-1 I I _ - - _ - - - - - i
I I In the calculations, 100 batches of 200 neutrons per batch were used, with three generations skipped to minimize the effects of the initial source distribution. The initial source was a flat distribution within a cuboid I enclosing the active core region and reflector regions. The k(eff) calculations are reported with an associated statistical uncertainty of one standard deviation (68% confidence level) that ranged from .0017 to .014 I with .002 as a typical value. Material Description I The materials and volumes used are included in Table D.4. Material volumes were calculated from information obtained from Fort Calhoun FSAR Tables; in the core volume inventory, as shown in Table D.3, water was assumed to I occupy the " unidentified" volume. The small differences in volume inventory between the calculations and the values quoted by FSAR could not be resolved, but these differences are expected to have a small effect on the resultant k(eff) of the crushed core. Since the calculated water volume is slightly higher, the removal of the water by replacing it with crushed solid volume components should be conservative or lead to a higher k(eff) than if less water were removed. No water displacers were assumed in the core. The greatest contribution to model inaccuracy is expected to result from the homogenization of the baron in the CEA's and in the shim rods, and in the smear-cut of the fuel. The inaccuracy associated with homogenization is considered to be slight and does not adversly affect the results of the analysis. Material I Zircolloy-4 composition was not available at the time calculations were made, so available Zircolloy-3 composition was assumed to sufficiently and accurately model the Zirc-4. The composition of the shim rods and control element assemblies was assumed to be that specified in FSAR. Conclusions and Compliance I These criticality analyses indicate that OPPD Fort Calhoun refueling operations in the Containment are in compliance with the NUREG 0612, Criterion II, Section 5.1 guideline for criticality potential due to the accidental drop of a heavy load (Reactor Vessel Head Assembly) onto a I refueled reactor core. The bounding worst-case potential is a refueled core containing 133 new 4 weight-percent enriched fuel assemblies. Criticality potential is postulated to be equal to or less for a partially refueled core and for a fully spent core. I I I D-2 I
I I APPENDIX D Table D.1 Specific Criteria II, NUREG-0612 Section 5.1 Initial Conditions / Assumptions Requested NUREG-0612 Item Requested Value Reference Source / Remarks I Attachment (3) 1.a Water /UO2 Vol Ratio 1.60 2.00 FSAR 11.2-11 KENO IV Code Cales 1.b Refueling water 1700 ppm Tech Spec. 2.8 1.c Amount of neutron 1.36 kg. Calcs from FSAR Fig. 3.4-1, poison in fuel Vol. I, Natural Boron 1.5 Fuel Enrichment 4 w/o Ref. E.4'and E.5 (worst case) 1.e Reactivity insertion k(eff)=+0.05 NUREG-0612 allowance value due to crushing of for maximum core. 1.f k(eff) allowed by 0.82 Calculated or derived from Technical Specs reactivity values given in during refueling FSAR Table 3.4-1, Vol. 1 l I I l I i l l I I I --
I I TABLE D.2 OPPD k(eff) for Crushed Core Resulting from Postulated Accident Drop of RV Head I 1. Uncrushed case (FSAR Vol 1) CEA's w/d & no H20-k(eff) 1.18 FSAR Vol I Table 3.4-1 Boron. With all boron in 0.82 Calculated shutdown condition (CEA's in, I 1700 ppm H2O) Change in k(eff) -0.36 I 2. Crushed case With all boron in shutdown condition 0.82 (CEA's in 1700 ppm H2O)
/.dd ke=+0.05 as +0.05 bounding by NUREG-I 0612 allowance Crushed max k(eff) 0.87 0.90
- 3. Maximum crush in Z- 0.59 0.90 direction using KENO IV Code to max k(eff)
(Ref A.4) I Intermediate crush using KENO IV code 0.50 0.90 I I
g TABLE D.3 E OPPD FT. CALHOUN Core Volume Accounting For Major Components With Ratio Comparisons For FSAR Values I Calculated Accumulated KENO IV FSAR (1) Velume Volume Volume Volume Component (cm ) (cm ) Fraction Fraction I UO 2 0.55051 + 7 0.55051 + 7 0.2949 0.341 Zirc-4 Clad 0.18043 + 7 0.73094 + 7 0.09667 0.101 CEA's in GT 0.036227 + 7 0.76721 + 7 0.01943 ---- (49) Water (H 2O) 1.09953 + 7 1.86674 + 7 0.5890 0.548 I . Total 1.86674 + 7 1.86674 + 7 1.0000 0.990 a Ratios g l H2 0/ Total 0.5890 0.548 l Non-H 2 0/ Total Assume Total = 1.0 for 0.4110 0.452 Zirc-4/ Total FSAR values 0.09667 0.101 H20/UO 2 1.997 1.607 Notes: 1. FSAR Vol. 4, p11.2-11 for average fuel assembly. I I I I
I I TABLE D.4 OPPD FORT CALHOUN Core Volume and Mass Molecular and Atomic Components for Active Core Materials Core Volume Mass Component (cm ) (gm) Uncrushed Core 1.86674 + 7 ---- UO 2 5.5051 + 6 5.5585 + 7 Uranium 4.8996 + 7 U-238 4.7036 + 7 U-235 1.9598 + 6 I 0-16 (oxygen) 6.589 + 6 Water - H 2O 1.09953 + 7 1.09953 + 7 H-1 (hydrogen) 1.2304 + 6 0-16 (oxygen) 9.7649 + 6 B-nat in H O2 (1700 ppm) 3.24 + 3 B-nat in shims 1.36 + 3 B-nat in CEAs 3.992 + 5 Alumina in shims 7.861 + 4 3.121 + 5 I A1-27 0-16 Zircolloy Clad 1.8043 + 7 1.057 + 5 9.2 + 2 1.1728 + 7 I Zr (0.9943)* Fe (0.0025) Ni (0.0005) 1.1661 + 7 2.932 + 4 5.864 + 3 Inconel-CEA Clad 4.623 + 5 I 5.497 + 4 Ni (0.76) 3.513 + 5 Cr (0.158) 7.30 + 4 Fe (0.072) 3.33 + 4 I Mn (0.002) Si (0.002) Cu (0.001) 9.0 + 2 9.0 + 2 5.0 + 2 I B4C-Carbon (shims & CEAs) Partially Crushed Core 2.22 + 3 All same except H 0 2 I H ,0 1.662 + 6 contributions 1.662 + 6 H 1.8598 ' 5 I O B-nat (H 2O) 1.4760 + 6 4.90 + 2 Fully Crushed Core All same except H O 2 contributions H,,0 0 0 H 0 0 0 0 0 B-nat (H 2O) 0 0 I
- Numbers in () are weight percent fractions.
M M M M M M M M M M M M M M M M M M M TABLE D.5 KENO IV Codo Hixing labios for OPPD TORT CAltlOUN Activo Coro Atom Densities (#Atojn.1) and KENO Fuel Hosonanco 10 ba rn. cm Hix httio ID No. Component No. Number Uncrushed Core Partly crushed Coro f ully crushed Coro i 1 11 - 1 1 1102 3.9379-2 1.1905-2 0 2 B-nat 1 5100 1.2050-3 2.3935-3 2.9083-3 3 C-12 1 6100 5.963-6 1.193-5 1.451-5 4 0-16 1 8100 3.2978-2 3.2527-2 3.2330-2 5 Al-27 1 13100 1.2638-14 2.5276-86 3.0750-14 6 Si 1 18:100 1.036-6 2.068-6 2.516-6 7 Cr i 24100 4.529-5 9.058-5 1.102-4 8 Hn 1 25100 5.28-7 1.056-6 1.285-6 9 Fo 1 26100 3.617-5 7.234-5 8.801-5 10 Ni 1 28100 1.9626-4 3.925-14 4.7753-14 11 Cu 1 29100 Rosonance 2.514-7 Resonanco 5.08-7 Rosonanco 6.178-6 12 Zr 1 10100 4 EDio ID 84 , 121 0-3 KFNO ID 8.2880-3 1 fi[N0_1D 1.0034-2 13 U-235 1 Varios -92509 0.0659-4 -92508 4.7805-4 -92507 6.4893-4 4 18: u-235 1 With -92510 2.6240-4 -92509 0.5993-4 -92508 0.0956-84 15 u-23 t1 1 C rush 92824 4.2293-3 92808 7.90f ils-3 92808s 1.0680-2 4 16 U-238 1 92825 2. 18:50-3 92809 4. 88:146 -3 92805 0.81870-2
I I TABLE D.6 KENO IV Code Fuel Scattering Resonance Corrections j UncrusFed Partial Crush Full Crush Component 9 c; f c', f ff U-238 12.0 7.6492-2 1.5299-1 1.8612-1 U-235 10.0 2.6399-3 5.3798-3 6.5449-3 I H 21.0 8.2696-1 2.5000-1 0 0 3.7 1.2202-1 1.2035-1 1.1962-1 3-nat 4.0 4.8200-3 9.5740-3 1.1633-2 A1 1.4 1.7693-4 3.5386-4 4.3050-4 E Zr 6.2 2.5569-2 .. 5.1138-2 6.2211-2 Fe 11.4 4.1234-4 8.247-4 1.0033-3 I Ni Cr Mn 17.5 4.3 2.1 3.4346-3 1.9475-4 1.11-6 6.8691-3 3.895-4 2.22-6 8.3568-3 4.739-4 2.70-6 Si 2.2 2.27-6 4.55-6 5.54-6 Cu 7.5 1.90-6 3.81-6 4.634-5 C-12 4.6 2.743-5 5.486-5 6.674-5 I s(core) 1.0628 0.5979 0.3965 U-238 : C'p = 166.73 46.90 25.57 U-235 : C'p = 3951.08 1111.44 605.84 Potential scattering cross sections of core are calculated by: 7 s (core) = E 0,g fi I where O'd f th element or component Os th fi = atom density for i element and C'p = E s(core) fi where O'p lies on one or proportioned between two values in KENO Table of O'p vs ID for U-238 and U-235. I I I I -
I I FIGURE D.1 REGION'4ISE MODEL REACTOR FOR XENO IV CODE I ,- _ ~~, s, I ,' Cyl H 2O s x
's is ,
i, - Reflector Above Core s
,I I
s , - ,
' '~ -- -
1-
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I l t i i i i I a ! I
, Uniform 'I Y Crush 8 l I !
I l u 1
~j 9,
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Active Core g , Region u l , e ( ' h Shroud 7 i 1 I l l l
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Y I e , Origin s e of - X or y w ' I Coord " i '
, 1 (0,0,0) i i i
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- Cyl H 2 O ' , l, ' 3 ~
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*=m Reflector Below Core ,
i Cuboid Box Enclosure / 6 g n, e.g1o. =b.r f Reflector is approximated by 1 ft thick water envelope around the core surface lI - I
il
!I i
FIGURE D.2 REACTOR DDIENSIONS FOR KENO IV CODE MODEL I
- m G
_ _u .. . a i E k C 1x* Top Reflector Circumferential
@og A - Reflector r
a,,,-~~~~~, , Region
!( -
) I g _ _ _ _ _ - ; ;
C3 . I : A C C e A. Active Core ! 1
@e Q C. Equiv Dia = 106.448" ! '
- 4Al C m M I @
m s a Ms
- iM C
N m Q m Si 3O b : Active length = 128" I @ M m U C -Q Ui m
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O G wa
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4 O g,4 O C : 4 u 9 r% q O *M A M M N CO M
.ig 3
I C #C C/) .C C.4 @ U Q 3 c C 4 e .-4 .c 3 0 .c
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43 CJ
- D W w am W QJ i G su QJ O G
- u. m un e ut u u sa u W .: C; O M O l j >i s-. :2, u .e vi I s
~~~- _i -
I i ' _ f Bottom Reflector i , , i, I (ralius) j l
~
C! : : 1 i k i e C C mj : l t C! C C NI N. mi e I CC C. C C: -i -
- che N m., W M; C..' O. M.'
. C. l N. . ,
e: C' c M .I l Ct M MI l s. e e n!
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l l ' l -! hs. ClJ - I> z'
- I I
I ll ( , x - 256 SHIMS, 0.3815 in. POISON 0.D.,1.10 w/o B 4C IN Al 023 m - 192 SHIMS, 0.300 in. POISON 0.D., 2.62 w/o B4C IN Al 023 ll *- 16 WATER DISPLACERS, 0.985 in. CLAD 0.D., Al 0 IN 0.100 in. THICK ZlRCALOY TUBING 23 h
- I o -e--CORE LOCATION NUMBER oF1o oH10
;! o --e- -FUEL ASSEMBLY T.YPE oCp x 4
oC2o oD2o oE2o oG20 *ox 2g o o o o l oC0 0 OC o oB 0 o x x 1x x oA00 x 0 IOx oB3o *gC *o D oE3o *oG oJ3 o
,l o
oC0 0 x0 0 x0 1x x o 0 OA 0 x OAO-O o B 1x 0 x oB4 o* 4 o*. oD4o "gE4,o" oG40 "o}4,o" o o o o o ll oCyx o* x 0 Ox OA0 o B 0, OA0 0 [B'2S oB5a oC5 o "O,05g 'oE5 o "o,G5,o" oJ5o l oA6o oB0 o 0 o oA00 [B , A0 o 2 o 0 o 8B"poa OA0 0 o
$ oC0 0 0 0 Y:o oD7o "o,Ug oG7o "o,J7,o" o o o o o"o, l O A8[
o 0 OA0 0 x x Bfox oA0o Bg oA 0 0 gBp, oCf i *o,B 9 a oC9o "o,090" oE9o "o,G 9,o" oJ90 l ( o o oA 0 0 88'" o o OA0 0 o [B'2, o OA0 0 o [B 1x 0 2 l Nu[p!c$ent ::0. a o g Location of Boron Poison Rods por afh n tat $n Un t 1 3.4-1 l . _ _ .. _. . _ ... _ _ . . . _ ____ ___
1 - - - _ _ . . _ - , , - - - !I
- I II lI I
lE APPENDIX E RADIOLOGICAL DATA !I . 4 I i ig ' 4I i 1 , ll 1 4 i
'I il 1 ;I
- I APPENDIX E RADIOLOGICAL ANALYSIS I The purpose of Appendix E is to present the conclusions and to describe the methodology utilized in conducting the analysis. It also establishes the conservatism of the assumptions utilized for determining the postulated I radiological release consequences of a heavy load drop onto the Fort Calhoun spent reactor core during refueling operations.
- 1. Conclusions
- a. During refueling operations, the Fort Calhoun Containment can E succe==fu127 miti8 ** a postulated radiological release frem an 3 accidental crush of a full core of spent fuel. Under this type of operation, the Containment is capable of maintaining site releases within NUREG 0612 guidelines of one-fourth 10CFR Part 100 limits for spent fuel cooled to a minimum time of 72 hrs.
- b. Trend analyses indicated the following:
- 1. For the postulated whole core damage scenerio; (72 hr cooling) containment isolation is required during refueling operations to maintain potential site radiological releases within the limit specified by NUREG 0612. Iodine filtration I. alone would be insufficient for both 72 hrs cooling and 240 hrs cooling (normal operations) due to the noble gas whole body dose contribution.
- 2. The Containment closed leak rate is approximately a factor of _
2 (conservative) for containing whole core releases from 72 I hrs cooled fuel and approximately a factor of 4 (conservative) for 240 hrs cooled fuel. (i.e., at the closed leak rate, 2 whole core releases could be contained for 72 brs cooled fuel and 4 whole core releases at 240 hrs cooled I fuel).
- 3. The number of fuel assemblies permitted to be damaged varies I .
inversely with containment leak rate, and directly with fuel cooling time within the limits analyzed. I
- c. Source terms for gaseous fission products available for release from spent PWR fuel required by NUREG 0612 and Reg. Guide 1.25 are about a factor of 10 (conservative) with respect to actual source terms available.
- d. Gaseous fission product source terms used as generated by the DRAGON (Ref. A.3) computer code are conservative with respect to source term values provided by OPPD. DRAGON generated activities were, on the average, at least 8*. higher than those values I provided by the District.
I E-1
/
I
I
- e. Small increases in potential whole body release doses of up to about 25% due to Xenon activity increases would be expected for higher weight fuel loads (four weight percent enrichments). These
.I expected increases are still well within the NUREG 0612 conservative credit allowances, and well within the factor of 2 (conservative) limit for closed containment.
- 2. Methodology Case A. Uncontained, unfiltered and iodine-filtered source terms for radiological releases resulting from whole core damage.
- 1. Generate gaseous fission product whole-core inventory activities at scram using DRAGON code with the following inputs:
Reactor power : 1500 Mwt Days full power: 432 Uranium in core: 48996 kg (FSAR Table 3.4-1) 235 Fraction U : 0.02235 (Weighted avg from FSAR)
- 2. Compare the source term activities provided by, OPPD with those generated by DRAGON and select the most conservative I values. Additional DRAGON program runs were made at different U235 w/o enrichments, 432 full power days, and 1500 Mwt. These analysis indicated a small increase in Xenon (up to 25%) and a very small decrease in Krypton activities.
,l These variations are well within the NUREG 0612 conservative 5 release fractions used.
I It should be noted that the particulate activity (i.e. all other nuclide contributions) was assumed negligible in the gas released. In addition, the radioactive bromines were assumed decayed out at the minimum fuel cooling time of 72 hrs.
- 3. Decay the DRAGON generated activities to cooling times of 72 hrs, 240 hrs, and 1000 hrs by RADIOISOTOPE Computer Code (Ref
.I A.1).
- 4. Reduce the decayed core activities by allowable NUREG 0612 credits (Reg. Guide 1.25).
Peaking factor (whole core): 1.2 Iodine available for release: 0.1 all except Il29 use 0.3 Noble gas available for release: 0.1 all but Kr 85 use 0.3
. = Iodine water reduction: 0.01 ;I .I E-2 I
I
I By utilizing the methodolgy presented in item 4, the values presented in Table E.6 were obtained. To determine the conservative extent of these NUREG 0612 credits with respect I to iodine and noble gases available for release, the fractions in Table E.7, the Indian Point FSAR and the Westinghouse RESAR were taken as typical PWR gas fractions.
- 5. Develope X/Q values for Exclusion Zone Boundary (EZB) and Low Population Zone (LPZ) using actual site meteorological data.
I 6. Estimate EZB and LPZ, 0-2 hr potential radiological release doses by inputting the reduced activities (from each set of 4 above) and X/Q into the REM 123 computer code (Ref. A.2). 7. Alternatively reduce iodine activities by the factor 0.05 to simulate a filtering effect and run PIM123 code. Results of steps 6 and 7 are shown in Table E.8 and Figure E.2.
- 8. Since the release doses postulated for full core failure exceeded the limits established by NUREG 0612 the number of
- g assemblies which could be damagad and still comply with said
.g limits was determined (Table E.9). For these, analyses it was assu=ed that the total release is a multiple of the single assemblies contained in the core. These values are presented on Fig. E.3, which also depicts the values defined by Table 2.1-2 of NUREG 0612. The two sets of values were compared to determine the conservatism of l the approach. . Casa B Contained and Icdine-recirculation filtered source terms for radiological releases resulting from whole core damage.
- 1. The DRAGON code's site and containment release rate options were utilized to estimate doses inside the containment and I those released to the environment. For these release rate integrated doses (30-day) that exceeded one-fourth 10CFR Part 100 site limits, the number of assemblies permitted to be g damaged and remain within the limits established by the NCPIG E were determined. Noble gas activities input were the decayed activities at appropriate cool times reduced by 0.12 factor allowed by NUREG 0612 or Reg Guide 1.25. Iodine activities
.I were the water-filtered (x 0.01) decayed activities. DRAGON code further reduced these activities by the recirculation filter removal fraction.
- 3. Compliance The radiological analyses conducted indicate that the OPPD Fort Calhoun
'I refueling operations in Containment are in compliance with NUREG 0612, Criteria I, Section 5.1 guidelines for a potential radiological release due to the accidental drop of a heavy load from the Polar Crane onto a spent reactor core.
E-3 I I
I I TABLE E.1 SPECIFIC CRITERIA I, NUREG-0612 SECTION 5.1, INFORMATION AND ASSUMPTIONS AS REQUESTED BY A'ITACHMENT 2 NUREG-0612 Item Requested Value Reference Source / Remarks l Attachment (2) 1.a Time after shutdown 72 hrs min OPPD FSAR Vol 5, p 14.18-2b. (Also Appendix A.1.2) 20 hrs typ OPPD -verbal communication I No. Fuel Assemblies draaged 13 *a (Full Core) OPPD FSAR VOL 1, Table 3.6.1 Assume maximum case Assumed duration 0-2 hrs Initial reless; int, containment of radiological 0-30 days Containment atmosphere ce site release atmosphere I 1.b Table 2.1-2 Items (Appendix A.l.5) Reactor Power Level 1500 Mwt(max) Operating License Limit 0-2 hr X/Q=EZB 1.54 - 3 see Calc 5% worst meterorological 3 m conditions 1 0-2 hr X/Q-LPZ 8.96 - 5 sec Calc 5% worst meteoro1gical 3 m conditions Peaking Factor 1.2 For full core, assume NUREG-0612 l5 Table 2.1-2 value No. fuel assemblies 133 See 2nd item above 'I in core , Refueling water 100 Reg. Guide 1.125 standard decon factor (conservative) all iodines Filter efficiency Elemental 95% Assume NUREG-0612 values. Organic 95% Actually conservative wrt. site I values of 9 9".-FS AR Cooling time 72 hrs See item 1 above 1.c Site-Specific Table E.3 See basis column' Table E.3 Other A sumptions 1 I
I l TABLE E.1 (Continued) SPECIFIC CRITERIA I, NUREG-0612 SECTION 5.1, I INFORMATION AND ASSUMPTIONS AS REQUESTED BY ATTACHMENT 2 NUREG-0612 Item Requested Value Reference Source / Remarks 1.d Rapid containment I isolation time (details continued) 0 Assume closed by tech spec on refueling. By FSAR Vol 3, Sec 7 and Fig 6.2-5 max time delay 5 see to close purge system 1.d.1 Detector Location -- FSAR Vol 3, Sec. 7 and Sec. 11 and Isolation Response Times for I Containment Isolation Purge valves 3 sec From instant of detection of I receive signal Valve Closure 2 see high-radiation signal From instant closure signal is initiated until closed. 1.d.2 Engineered Safety -- Associated instrumentation and I Features (EST) controls designed and built as class 1 to IEEE 279-1968 code. 1.d.3 ESF System - FSAR Section 6.4, " Containment Compliance with Air Recirculation, Cooling and Reg. Guide 1.52 Iodine Removal System"; Section 7.3, " Engineered Safeguards I Controls" I 1.d.4 Operational & maintenance ESF Normal plant operation, all ESF maintained in ready condition, and containment closed. Refueling operations, containment I purge exhaust fans low vol rate and discharge through filters to stack where air is monitered. Upon monitor alarm, containment I isolated; FSAR 7.3.2.6 Isolation closure of cent. purge 2 sec Containment closed Required by Technical I " Refueling during refueling Specification 2.8 Operations" I I I
I 1 I i l TABLE E.2 CALCUISTED OPPD SITE DOSES FOR DIFFERENT CONTAINMENT LEAK RATES AND FUEL COOLING % TIMES CONSEQUENTAL TO REACTOR VESSEL HEAD DROP ACCIDENT (Results from DRAGON Computer Code, Ref. A.3) Table Values in REM Item Time Interval After RV Head Droo 0-2 hrs 0-8 hrs 0-24 hrs 0-4 da 0-30 da 72-hr Cooled Fuel Closed Containment (0.0014) Containment Thyroid 0 0 0 0 0 Containment Whole Body 1.4+5 5.41+5 1.52+6 5.00+6 1.23+7 Site Thyroid 0.718 0.718 0.718 0.718 0.718 Site Whole Body 0.164 0.633 1.30 2.13 3.21 I 2 x closed leak rate Containment whole body Site Thyroid 1.40+5 1.44 5.41+5 1.44 1.52+6 1.44 4.99+6 1.44 1.23+7 1.44 Site Whole Body 0.328 1.27 2.59 4.25 6.42 4 x closed leak rate Containment whole body 1.40+5 5.41+5 1.52+6 4.98+6 1.21+7 I Site Thyroid Site Whole Body 2.87 0.66 2.87 2.53 2.87 5.19 2.87 8.48 2.87 12.73 10 x closed leak rate I Containment whole body 1.40+5 5.40+5 1.51+6 4.92+6 1.17+7 Site Thyroid 7.18 7.18 7.18 7.18 7.18 Site Whole Body 1.64 6.32 12.93 21.0 31.2 240-hr Cooled Fuel Closed Containment I Containment Whole Body 5.42+4 Site Thyroid Site Whole Body
.36-0.063 2.12+5 0.36 0.25 6.12+5 0.36 2.04+6 0.36 5.24+6 0.36 0.52 0.86 1.32 2 x Closed Leak Rate Containment Whole Body 5.42+4 2.12+5 6.11+5 2.04v6 5.21+6 Site Thyroid 0.722 0.722 0.722 0.722 0.722 Site Whole Body 0.126 0.495 1.03 1.71 2.63 I
I
~
I I
I TABLE E.2 (Continued) 4 x Closed Leak Rate I Containment Whole Body 5.42+4 2.12+5 6.10+5 2.03+6 5.15+6 Site Thyroid 1.44 1.44 1.44 1.44 1.44 Site Whole Body 0.25 0.99 2.07 3.41 5.22 5 10 x Closed Leak Rate Containment Whole Body 5.41+4 2.12+5 6.08+5 2.01+6 4.97+6 j Site Thyroid 3.61 3.61 3.61 3.61 3.61 3 Site Whole Body 0.63 2.47 5.14 8.47 12.76 NOTE: Doses are integrated from accident time to end of interval specified. I . 1/4 10 CFR Part 100 Limits: Thyroid = 75 rem; whole Body = 6.25 rem at site boundary. Inputs to code are in Table E.3. I I I I I I I I I I I I I
e I TABLE E.3 ASSUMPTIONS USED IN CALCULATING OPPD SITE DOSES FOR CONTAINMENT LEAK RATES I AND FUEL COOLING TIMES (INPUTS TO DRAGON COMPUTER CODE, REF. A.3) Values Parameter / Time Interval After RV Head Dron Basis and I Assumption 0-2 hrs 0-8 hrs 8-24 hrs 1-4da 4-30da Direction Error X/Q (EZB) 1.54-3 1.54-3 8.96-4 6.32-4 3.98-4 Calculated 5% (sec/m ) meteor. Site I Breath Rates (m /sec) 3.47-4 3.47-4 1.75-4 2.32-4 2.32-4 specific values Standard values Containment 1.40-3 1.40-3 1.40-3 7.00-4 7.00-4 FSAR values Leak Rates (Closed containment or reference values) I Standard values (Vols/24 hrs) Fractions of Noble gases: 1.2-1 Reg Guide 1.'5 & FP Gas Release Methyl halogens: 1.2-3 NUREG-0612 to Containment Elemental halogens: 1.2-3 Slightly I conservative I Chem spray credit None or 0 Conservative Assumption 3 Containment 1.05 + 6 ft FSAR Standard I Volume Value Containment 86500 cfm FSAR Vol 3, Tab. I Air Recire & Filter Rate (one only operational) 6.4-1,5,6 Conservative Recire Purify Elemental halogens: 0.95 Assume NUREG 0612 I Factors Meth fl halogens: 0.95 Conservative wrt
- actual of 99% -
Exhaust Air None or 0. Rate Assume exhaust closed. Standard on Cont. Leakage assumed in I Exhaust Air containment leak rates. N/A isolation Standard on Cont. Purification isolation. Transit Time None or O. Slightly from Source I conservative to Site
'Since all values are standard or conservative, the net effect is conservative.
I I
m mm mm mM M W m M M M M M M M M M TABLE E.4 OPPD SITE DOSES VS CONTAINHENT LEAK RATE ANO FUEL COOL TlHES f ull CORE REl. EASE OF CASEOUS FP (DRAGON CODE) 72-hr Cooled fuel 240-hr Cooled fuel Changed Release Release Containment 30-day Units REM Exceeds 30-day Units REM Excoods Condition EZD # Assemblies Limit 1/86 10 CFR EZD # Assemblies Limit 1/34 10CFit Pa rt 100 Pa rt 100 Thy WB Thy =75 Wu=6.25 Full Core Thy WB Thy =75 WD=6.25 Fu l l co re Rer: Closed containment 0.72 3.21 13854 258 N 0.36 1.32 27/08 629 N leak Rato .00184 Rocirc ril 86500 cra (1) Pa rt closed contain 1.88 4 6.82 4 6927 129 N 0.72 2.63 138584 316 N 2 x clos IK rate (ma rg ina l ) Roc i rc (2) Part ci sod contain 2.87 12.73 3475 65 Y 1. le te' 5.22 6921 159 N 86 x clos LK Rate Rocirc (3) Part closed contain 7.18 31.2 1389 26 Y 3.61 12.76 2763 65 Y 10 x clos LK rato 110 circ (8) OPIN containment 4 30000. 723 0.3 1.1 Y 15000. 3 ta 3. 0.6 2.4 Y No filter (0-2 hr) (0-2 hr) No recirc (REH123 Codo) (5) OPEN containment 1500. 6 784 6.6 1.2 Y 752. 331. 13.2 2.5 Y with 1-Filters (0-2 hr) (0-2 hr) (REH123 Codo)
I TABLE E.5 COMPARISONS OF GASEQUS FISSION PRODUCT l
=
ACTIVITIES AS PROVIDED BY OPPD WITH THOSE GENERATED BY DRAGON COMPUTER CODE FP OPPD DRAGON 3 DRAGON Nuclide Fort Calhoun Data Generated Data Conservative (Curies) (Curies) wrt OPPD I 129
- 1.04+00 8.97-01 131
- 4.15+07 3.57+07 I 132
- 133
- 134 6.03+07 8.43+07 9.10+07 5.19+07 8.50+07 9.56+07 135
- 7.84+07 8.06+07 136 -----
3.68+07 Sum 3.55+08 3.856+08 +8.47*. initial Kr83m ----- 6.67+06 85m* 1.07+07 1.65+07 85
- 3.38+05 6.18+05 I 87 88 89 1.96+07 2.80+07 3.45+07 3.14+07 4.50+07 5.88+07 Sum 9.314+07 1.590+08 +70.7*. initial Xe 131m* 2.93+05 2.13+05 I 133m*
133
- 135m*
8.43+07 1.69+07 2.06+06 8.48+07 2.56+07 135
- 1.47+07 I 137 138 7.42+07 6.78+07 1.17+07 7.70+07 7.89+07 r Sum 2.582+08 2.803+08 +8.56". initial
* = Significant activity after 72 hrs; Bromine nuclides activity essentially gone in 72 hrs, no bromines listed.
lI l I I I I
I TABLE E.6 REDUCED FISSION PRODUCT GAP ACTIVITIES BY NUREG 0612 CREDITS: GAS ACTIVITIES RELEASED THROUGH
- CONTAINMENT REFUELING WATER Curies FP Unreleased RG 1.25 Curies Released at Cooling Times
-I Nuclide 0 hr Fraction 72 hrs 240 hrs 1000 hrs I 129 0.897 3.6-3 3.23-3 3.23-3 3.23-3 131 3.57+7 1.2-3 3.40+4 1.87+4 1.24+3 132 5.19+7 1.2-3 3.86+4 8.62+3 9.74+0 133 8.50+7 1.2-3 9.25+3 3.42+1 3.4-10 -I 134 9.56+7 1.2-3 0 0 0 135 8.06+7 1.2-3 5.54+1 0 0 136 3.68+7 1.2-3 0 0 0 1 Kr83m 6.67+6 0.120 3.38-3 0 0 85m 1.65+7 0.120 2.33+1 0 0 85 6.18+5 0.360 2.22+5 2.22+5 2.21+5 8 87 3.14+7 0.120 0 0 0 88 4.50+7 0.120 9.23-2 0 0 89 5.88+7 0.120 0 0 I
0 90 0 0.120 0 0 0 91 0 0.120 0 0 0 I Xe131m 133m 133 2.13+5 2.06+6 8.48+7 0.120 0.120 0.120 6.29+5 1.45+5 8.06+6 1.25+6 1.86+4 3.31+6 5.62+5 1.12+0 5.21+4 135m 2.56+7 0.120 1.73+3 4.73-5 0 I 135 137 138 1.17+7 7.70+7 7.89+7 0.120 0.120 0.120 1.05+5 0 0 3.7-1 0 0 0 0 0 I 139 140 0 0 0.120 0.120 0 0 0 0 0 0 Notes: The curies released at cooling times were inputs to S&W REM 123 I Computer Code for unfiltered dose calculations. These values reduced by 0.05 to simulate 95% filter removal for Iodines only; Kr and Xe same on filtering. I I I I I 1 I
I TABLE E.7 ACTUAL CORE INVENTORY GAP AND PLENUM GAS FRACTIONS IN (W) RESAR AND INDIAN POIhT I PSAR PWR's AND THE CONSERVATIVE FRACTIONS ALLOWED BY NUREG 0612, REG. GUIDE 1.25 I FP Indian Point (1) RESAR (2) Reg. Guide 1.25 Nuclide PSAR Fraction Fraction Fraction I 129 ----- ----- 0.30 131 0.023 0.0188 0.10 132 0.0026 0.0021 I 133 134 135 0.0079 0.0016 0.0043 0.0063 0.0013 0.0036 0.10 0.10 0.10 0.10 Kr85m 0.0029 0.0029 0.10 85 0.2157 0.255 0.30 87 0.0020 0.0016 0.10 88 0.0029 0.0023 0.10 Xe131m ----- 0.0228 0.10 I 133m 133 135m 0.0127 0.0185 0.0086 0.0101 0.0153 0.00071 0.10 0.10 0.10 135 0.0054 0.0042 0.10 138 ----- 0.00074 0.10 Notes
- 1. Power level = 1000 Mwt, 1000 days
- 2. Power level = 2900 Mwt, 650 days References
, 1. Lamarsh, J. R., Introduction to Nuclear Engineering, Dec. 1977, Table 11.7, p. 547.
- 2. Westinghouse RESAR Rev 3, Table 15.1-4, p.15.1-26.
l il l !I l I I I
TABLE E.8 DOSES AT SITE EXCLUSION ZONE BOUNDARY (EZD) AND LOW POPULATION ZONE DOUNDARY ( LPZ) FOR INSTANTANEOUS Wil0LE-CORE RADIOLOGICAL ' RELEASES AT FUEL COOLING TlHES OF 72 itRS. 240 llR. AND 1000 llR UHFILTERED AND 80 DINE FILTERED (OPPD FORT CALHOUN) 72 hrs 240 hrg JDD u rg EZD LPZ EZD LPZ EZD 1.PZ Untiltered lhyroid Dose 3.00*4 1.74+3 1.50+4 8.75+2 9.81+2 5.71+1 whole Body Doso 7.23+2 4.21+1 3.43+2 2.00+1 5.31+1 3.13 lodino filtered Thyroid Doso 1.50+3 8.72+1 7.52+2 4.38+1 4.92+1 2.86 Whole Dody Dose 6.74+2 3.92+1 3.31+2 1.93+1 5.35+1 3.11 Results of REM 123 Computer Codo Runs , M/Q (EZD) = 1.54-3 sec/m3 X/Q (LPZ) = 8.96-5 sec/m3 ! (0-2hr) (0-8 hrs) 1
M M M M M M M M M M M M M M M M M M M TABLE E.9 NUMBER Of IUEL ASSEHDLIES DAMAGED TO HAKE 1/as 10 CFR PART 100 SITE LIMITS RADIOLOGICAL RELEASE UNCONTAINED FOR FILTERED AND UNFILTEllED 0-211R RELEASE 72 hr 210 4 hr 1000 hr fuel Cool Fuol Cool fuel Cool Unfiltored Ihyroid EZu 0.33 0.66 10.2 Whole Body EZD 1.15 2.42 15.4 l lodine fi t tored 1hyroid EZB 6.65 13.26 202.7 Whole Body EZD 1.23 2.51 15.5 Number assemblies = 133 f 1/2s IMS Part 100 limit dosel Total Duse Note: Unfi l te red is limited by lodino on thyroid doso; flitored is whole body limit rt. I 1
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! 'l' i I IC'* ' C'A i0 " l0" L IO Initial Containment Leak Rate (Containnent Vols/ Day)
FIGURE E.1 OPPD CONTAINMENT AFFECT ON THE LIMITING NUMBER OF WEL ASSD3 LIES DAMAGED FOR POTENTIAL SITE RADIOLOGICAL RELEASE TO REMAIN WITHIN 1/4 10CFR PART 100 LIMITS AT FUEL COOLING TIMES 72 HRS AND 240 HRS Results of DRAGON Code Runs; X/Q (FIB): Calc. site specific Limiting number fuel asse=blies, n = 133 1/4 10CFR Part 100 limit potential EZB dose I I
I .
, i i I \l i
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to l '! 20
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IIc ILD Days Fuel Cooling Time FIGURE E.2 , OPPD UNCONTAINED POTENTIAL RADIOLOGICAL RELEASE DOSES FOR WHOLE CORE DAMAGE AS FUNCTIONS OF FUEL COOLING TIME AT DAMAGE INDICATING MIND!UM COOLING TDIE TO MEET 1/4 10CFR PART 100 GUIDELINES i REM 123 Code results; X/Q: 1.54 (10-3) sec/m3 0-2 hr EZ3 i 8.96 (10-5) sec/m3 0-8 hr LPZ l lI I
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0,1 ' 0 to 2.0 30 40 50 GO TO TO 90 100 stC ILO Days Fuel Cooling Time I __ -- FIGURE E.3 NUMBER OF OPPD FUEL ASSEMBLIES THAT COULD BE DAMAGED TO REACH 1/4 10CFR PART 100 LIMITS US FUEL COOLING TIME Comparison With NUREG 0612, Figure 2.1-1 (PWR) I Results of R M 23 Code Runs OPPD: X/Q (EZB) = 1.54 (10-b)Uncontained 0-2 hr sec/m3 ; Power Level = 1500 Potential Mwt Releases I I I
I Criteria I & II, Appendix D, E References
.I A. Computer Codes A.1 RADI0 ISOTOPE; a qualified and propietary Stone & Webster Engineering Corp. code for calculating activities of radioisotopes I
as a function of decay time; September 1981. A.2 REM 123; a qualified and proprietary Stone and Webster Engineering Corp. code for coverting radiological releases to doses; April 1981. A.3 DRAGON; a qualified and proprietary Stone and Webster Engineering Corp. code for calculating dose and radioactivity from nuclear facility gaseous outflows; February 1977. A.4 KENO IV; "An Improved Monte Carlo Criticality Program"; L. M. Petrie, N.F. Cross; ORNL-4938; November 1975. A.5 Oh, Inki and Rothe, R.E.; "A Calculational Study of Benchmark Critical Experiments on High-Enriched Uranyl Nitrate Solution Systems"; Nuclear Technology, 4J , 226 (1978). B. Regulatory Guides _ B.1 RG-1.25 (SG-25) " Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors"; March 1972. B.2 RG-1.145, " Atmospheric Dispersion Models for Potential Accident I C. Consequence Assessments at Nuclear Power Plants"; August 1979. NUREG's I C.1 NUREG 0612 Control of Heavy Loads at Nuclear Power Plants: Resolution of Generic Technical Activity A-36; July 1980. C.2 NUREG 0554 Single Failure Proof Cranes for Nuclear Power Plants; May 1979. C.3 NUREG 0578 TMI-2 Lessons Learned Task Force: Status Report and Short-Term Recommendations; July 1979. C.4 NUREG 0737 Clarification of TMI Action Plan Requirements; November 1980. C.5 NUREG 75/087 Standard Review Plan for the Review of Safety ,I Analysis Reports for Nuclear Power Plants - LWR Edition; December 1975 and Revisions. I
il l l l D. Stone & Webstar I D.1 RP-13, Radiation Protection Manual for Pressurized Water Reactors; Stone & Webster Engineering Corp. ; June 1975. D.2 PWR Reference Nuclear Power Plant, Stone & Webster Engineering Corp.; as amended. E. Omaha Public Power District (OPPD) E.1 OPPD, Final Safety Analysis Report, Fort Calhoun Station, Unit No. 1; as amended. I E.2 OPPD, Operating License DPR-40 and Technical Specification, Fort Calhoun Station, Unit No. I with amendments and appendices. E.3 Fort Calhoun Station Unit #1, Environmental Reoort, section 5, I " Environmental Effects of Accidents", Stone & Webster Engineering Corp. from J. Monroe, OPPD; November 9,1981. I E.4 " Criticality Analysis for the Ft. Calhoun Nuclear Plant Spent Fuel Storage Racks", Pickard, Lowe and Garrick, Ine; July 1981; Stone & Webster Engineering Corp. from J. Monroe; November 2, 1981. I E.5 Amendment to Tech Spec 4.4.2, OPPD, Ft. Calhoun; Re: Design feature of racks for storage of new & spent fuel; Stone & Webster Engineering Corp. from J. Monroe; November 2,1981. I I I I I I I I I I
F. Miscellaneous ll 4
=
F.1 USNRC Memo: To all licensees of operating plants and applicants for operating licenses and holders of construction permits; December 22, 1980.
!3 F.2 USNRC Memo: To all licensees.
Subject:
Control of Heavy Loads 1 3 (GH 81-07); Febuary 3, 1981; Enclosure 3 missing pages. i lI 1 1. i lI lI lI lI I I I I I I _}}