ML20247P341
| ML20247P341 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 05/19/1998 |
| From: | Woodard J SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20247P348 | List: |
| References | |
| LCV-0828-D, LCV-828-D, NUDOCS 9805270418 | |
| Download: ML20247P341 (125) | |
Text
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.ey J. D.Woodard Southern Nuclear Executive Vice President Operating Company,Inc.
40 invemess Center Parkway Post Office Box 1295 Birmingham. Alabama 35201 Tel 205.992.5086 SOUTHERN h COMPANY Hsy 19, 1998 Energy to Serve YourWurid" Docket Nos. 50-424 and 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington. D. C. 20555
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Ladies and Gentlemen:
VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECHNICAL SPECIFICATIONS
. ADDITIONAL FUEL STORAGE RACKS FOR UNIT 1 FUEL STORAGE POOL REQUEST FOR ADDITIONAL INFORMATION By way ofletters LCV-0828-A and LCV-0828-B dated September 4,1997 and November 20,1997 respectively, Southem Nuclear Operating Company (SNC) requested to amend the Vogtle Electric Generating Plant (VEGP) Unit I and Unit 2 Technical Specifications, Appendix A to Operating Licenses NPF-68 and NPF-81. The revision to the Technical Specifications is to change the capacity of the Unit I spent fuel storage pool from 288 assemblies to 1476 assemblies, and to revise the Design Features
- description to reflect the criticality analyses and storage cell spacing.
Letter LCV-0828-A contained a report entitled " Modification Report for Increased Spent Fuel Pool Storage Capacity". The report addressed thermal hydraulic safety, seismic and structural adequacy, radiological compliance, and mechanicij integrity of the racks.
Letter LCV-0828-B submitted the results of the criticality analyses along with the formal request for the Technical Specification amendment.
This letter provides the SNC response to the request for additional information (RAl) f received on April 10,1998. It also includes responses to questions from the Staffin addition to the RAl.
At the request of the Staff, we have reviewed the SER dated June 16,1982 and f
. supplement dated October 22,1982 for the Maine Yankee Atomic Power Company N
(MYAPC) Docket No. 50-309 and concur that this SER is applicable to the spent fuel racks obtained from MYAPC for use at VEGP. The racks have not been modified for their use at VEGP.
9805270418 980519 PDR ADOCK 05000424 4
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(U. S. Nublear Regulatory Commission Page 2 of 4 LCV-0828-D
. During the preparation for the re-racking,it was discovered that the discussion of the J
heavy loads consideration as. stated in Section 3.5 of the modification report submitted to the NRC does not fully describe the re-racking process to % performed at VEGP.
Specifically, Section 3.5 states that the lin rigs are load tested with 300% of the maximum weight to be lined. This provision is derived from ANSI N14.6 which refers to the requirements for moving heavy loads over or near spent fuel assemblies or L
- unprotected safety-related equipment. As stated in both the modification report and the L
10 CFR 50.92 evaluation previously submitted, all spent fuel wi!! have been relocated to L
the. Unit 2 spent fuel pool, and mechanical stops will be installed to ensure that no heavy loads will be carried over spent fuel assemblies as a part of the Unit I re-racking effort.
SNC will load test the temporary gantry crane and all liA rigs at 125% of the maximum load to be handled in accordance with ASME B30.2 " Overhead and Gantry Cranes",
followed by an inspection of the critical welded joints. Safe load paths have been defined l-and liAing requirements for maintaining a maximum travel height, providing single-failure-prooflining, or providing redundant rigging have been designated for these areas as described in NUREG-0612. of this letter contains Chapter 3.0 of the modification report in its entirety i'
which contains the above-described revision to Section 3.5 of the report. SNC has reviewed the previously submitted 10 CFR 50.92 evaluation and has determined that the f
conclusion that no significant hazards will result from the proposed license amendments remains valid.
. The RAI transmitted to SNC on April 10,1998 contained three major groups of questions. The groups were: 1) radiological; 2) civil engineering and geoscience; and 3) thermal hydraulic and lining. The responses to these questions, by group, are included in Enclosures 2,3, and 4. Enclosure 5 contains the responses to additional questionr from the Staff.
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Portions of the responses to Questions 1,4: 6,11, and 13 of Enclosure 3 contain j
information proprietary to lloltec International. The proprietary responses were submitted under a separate cover by way ofletter LCV-0828-C. The proprietary portions of the L
. questions in this letter have been blacked out. Attachment 1 of Question 1 and i Attachments 1 and 2 of Question 11 are also proprietary and have been excluded from this letter.
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- Page 3 of 4 LCV-0828-D I
l SNC recommends that the commitments listed b.-low become license conditions in Appendix D of Facility Operating Licenses No. NPF-68 and NPF-81:
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- 1. The spent fuel pool heat loads will be managed by administrative controls. These controls will be placed in applicable procedures prior to transfening irradiated fuel l
into the Unit I spent fuel pool.
- 2. The spent fuel pool heat loads will be managed by administrative controls. These controls will be placed in applicable procedures prior to transferring additional irradiated fuel into the Unit 2 spent fuel pool.
- 3. The FSAR will be updated to include the heat load that will ensure the temperature limit of 170 F will no't be exceeded as well as the requirement to perform a heat load j
evaluation prior to transferring irradiated fuel to either pool. This will be included in i
the next appropriate FSAR update following the installation of the Unit I spent fuel l
racks.
Mr. J. D. Wocdard states that he is an executive vice president of Southem Nuclear and is authorized to execute this oath on behalf of Southem Nuclear and that, to the best of his knowledge and belief, the facts set forth in this letter and enclosures are true.
Sincerely,
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~A d D.\\ oodard Sworn to and subscribed before me l
this/
day of M o ],1998.
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l U. S. Nuclear Regulatory Commission Page 4 of 4 LCV-0828-D L
Enclosures:
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- 1. Revised Chapter 3.0 " Material r.nd Ileavy Load Considerations" J
- 2. Responses to RAl, Group 1 Questions. Radiological
- 3. Responses to RAI, Group 2 Questior.s, Civil Engineering and Geoscience j
- 4.. Responses to RAI, Group 3 Questions, Thermal liydraulic and Liftmg
- 5. Responses to Additional Questions xc:
Southern Nuclear Operating Company Mr. J. B. Beasley (w/o att.)
Mr. M. Sheibani (w/o att.)
NORMS U. S. Nuclear Regulatory Comrnission Mr. L. A. Reyes, Regional Administrator Mr. D. II. Jaffe, Senior Project Manager, NRR Mr. J. Zeiler, Senior Resident Inspeu or, Vogtle i
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LCV-0828-D w
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4 ENCLOSURE 1 MODIFICATION REPORT FOR INCREASED SPENT FUEL STORAGE j
i REVISED CIIAPTER 3.0
" MATERIAL AND IIEAVY LOAD CONSIDERATIONS" l
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3.0 MATERIAL AND HEAVY LOAD CONSIDERATIONS 3.1 Introduction Safe storage of nuclear fuel in Vogtle 1 requires that the materials utilized in the rack fabrication be of proven durability and be compatible with the pool water environment. All activities in the rerack installation process are required to comply with the provisions of NUREG-0612 to I
eliminate the potential ofinstallation accidents. This section provides the necessary information on these two subjects.
3.2 Rack Structural Materials All material used in the construction of the rack modules is series 300 stainless steel except for
'1 the pedestal support legs which are precipitation hardened A564-630 and the poison material which is discussed below.
3.3 Poison Material (Neutron Absorber)
In addition to the structural and non-structural stainless steel material, the racks employ Boral*,
a patented product of AAR Manufacturing, as the neutron absorber material. A brief description O
of Borai feiiewe.
Boral is a thermal neutron poison material composed of boron carbide and 1100 alloy aluminum.
Baron carbide is a compound having a high boron content in a physically stable and chemically inert form. The 1100 alloy aluminum is a lightweight metal with high tensile strength which is i
protected from corrosion by a highly resistant oxide film. The two materials, boron carbide and aluminum, are chemically compatible and ideally suited for long-term use in the radiation, thermal and chemical environment of a nuclear reactor or a spent fuel pool.
Boral has been exclusively used in fuel rack applications in recent years. Its use in the spent fuel pools as the neutron absorbing material can be attributed to its proven perfonnance (over 150 pool years of experience) and the following unique characteristics:
i.
The content and placement of boron carbide provides a very high removal cross-section for thermal neutrons.
ii.
Boron carbide, in the form of fine particles, is homogeneously dispersed throughout the central layer of the Boral panels.
iii.
The boron carbide and aluminum materials in Boral do not degrade as a result oflong-term exposure to radiation.
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The neutron absorbing central layer of Boral is clad with permanently borM surfaces of aluminum.
v.
Bo;al is stable, strong, durable, and corrosion resistant.
As indicated in Table 3.3.1, Boral has been licensed by the USNRC for use in numerous BWR and PWR spent fuel storage racks and has been extensively used in international nuclear installations.
Boral Material Characteristics Aluminum: Aluminum is a silvery-white, ductile metallic element. The 1100 alloy aluminum is used extensively in heat exchangers, pressure and storage tanks, chemical equipment, reflectors and sheet metal work, it has high resistance to corrosion in industrial and marine atmospheres. Aluminum has atomic number of 13, atomic weight of 26.98, specific gravity of 2.69 and valence of 3. The physical, mechanical and chemical properties of the 1100 alloy aluminum are listed in Tables 3.3.2 and 3.3.3.
The excellent corrosion resistance of the 1100 alloy aluminum is provided by the protective oxide film that develops on its surface from exposure to the atmosphere or water. This film prevents the loss of metal from general corrosion or pitting cerrosion.
Baron Carbide: The boron carbide contained in Boral is a fine granulated powder that conforms to ASTM C-750-74 nuclear grade Type II. Typical properties of baron carbide are provided in Table 3.3.4.
The rack modules to be installed in the Vogtle 1 pool come from the Maine Yankee fuel pool.
Thus, the Boral has already been passivated in a water environment during the time period the racks were in the Maine Yankee fuel pool. The protective layer of oxide film on the Boral panels assures that the integrity of the Boral will be maintained during the life expectancy of the Vogtle fuel racks.
3.4 Compatibility with Coolant All materials used in the construction of these flux trap racks have an established history ofin-pool usage. Their physical, chemical and radiological compatibility with the pool environment is an established fact at this time. As noted in Table 3.3.1, Boral has been successfully used in both vented and unvented configurations in fuel pools. Austenitic stainless steel is perhaps the most widely used stainless alloy in nuclear power plants.
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3.5 IIeavy Load Considerations for the Proposed Reracking Operation Safe load paths, meeting NUREG-0612 criteria, will be followed for moving the existing and new racks in the Fuel Handling and Auxiliary Buildings. The fuel racks will not be carried directly over any fuel. The areas defined in the safe load path will be reviewed for safe shutdown equipment beneath or directly adjacent to a potential travel path. These areas will be defined by a maximum travel height or single-failure-proof lifting or redundant rigging guidelines as described in NUREG-0612.
The Fuel llandling Building Bridge crane is currently rated for 125 tons at the main hoist, and contains a 15-ton auxiliary hook. The maximum weight for the racks to be installed is 25,075 lbs and that for the existing two racks is 33,000 lb=. The main hoist of this crane will be utilized to tmnsfer racks to and from the Fuel Building. As a result, this overhead crane is qualified to accept the anticipated load during the rerack project.
Within the Fuel Building itself, in the area of the SFP, there presently is no overhead crane to handle the rack movements. In order to handle the new rack installations and the removal of the existing racks, a temporary gantry crane with a hoist rated for 20 tons will be erected on the existing fuel bridge rails. This crane was used to set the existing racks. Mechanical stops wi i be in place to prevent this gantry crane from carrying lifL over on positioned over new or spent fuel.
Prior to the start of reracking operations, the spent fuel will be moved from the Unit 1 pool to
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the Unit 2 pool. This fuel movement and storage has been previously approved.
Lifting devices will be used to lift the empty existing and new rack modules. ~1nese lifting devices, including standard lift rigs and spreader bars, will be evaluated and shown to satisfy the evaluation criteria of NUREG-0612 Section 5.l(I - IV). The approach adopted for this project is in agreement with NUREG-0612 to provide assurance by utilizing safe load paths away from irradiated fuel, evaluating heavy load drops over safety-related equipment, and the use of dual load path rigging where absolutely required. Since these lifting devices are not required for remote engagement and will not adversely affect any safety-related system as defimed in ANSil4.6-1978, these devices are not required to meet the requirements of a "Special Lifting Device." Also, the ANSil4.6 defines a " critical load" as one which could " adversely affect any safety-system."
Pursuant to the defense-in-depth approach of NUREG-0612, the following additional measures of safety will be undertaken for the reracking operation.
i.
The crane used in the project will be given a preventive maintenance checkup and inspection per the Vogtle maintenance procedures before the beginning of the reracking operation.
ii.
The existing fuel racks will be lifted just above the pool floor and held at that elevation for a length of time before making any further rack movements.
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r iii.
Safe load paths will be followed for moving the existing and new racks in the Fuel Handling Building. Prior to the start of reracking o; trations, the spent fuel will be moved from the Unit 1 pool to the Unit 2 pool. Mechanical stops will be installed to ensure that the fuel racks can not be carried directly over or near any fuel located in the Unit 2 pool.
l iv.
The rack upending and lay down will be carried out in an area which has been l
reviewed for p;oximity to safety-related functions. This evaluation also includes l
the review of heavy load drops, single failure and redundant rigging criteria.
v.
All crew members involved in the use of the lifting and upending equipment will be given training.
vi.
All heavy loads are lifted in such a manner that the C.G. of the lift point is aligned with the C.G. of the load being lifted.
All phases of the reracking activity will be conducted in accordance with written procedures which will be reviewed and approved by SNC.
The proposed compliance with the objectives of NUREG-0612 follows the guidelines contained iu Section 5 of that document. The guidelin a of NUREG-0612 call for measures to " provide an adequate defense-in-depth for handling of heavy loads near spent fuel.. ". The NUREG-0612 guidelines cite four major causes ofload handling accidents, namely i.
operator errors ii.
rigging failure iii.
lack of adequate inspection iv, inadequate procedures The Vogtle 1 rerack program ensures maximum emphasis on mitigating the potential load drop accidents by implementing measure.c to eliminate shortcomings in all aspects of the operation including the four aforementioned areas. A summary of the measures specifically planned to deal with the major causes is provided below.
Operator errors: As mentioned above, Southem Nuclear Operating Company plans to pzovide comprehensive training to the installation crew.
I Rigging failure: The lifling devices designed for handling and installation / removal of the racks j
at Vogtle 1 will be analyzed for and designed to the maximum lift required during the rerack l
operation. Each lifting device will be load tested per the requirements of NUREG-0612 and its appropriately referenced standards including ASME B30.2. All lift heights will be analyzed such that an uncontrolled lowering will not adversely impact safe shutdown equipment. The load path will not be over or near irradiated fuel in the Unit 2 spent fuel pool.
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The lift rig for the new rack modules is the same rig used by Maine Yankee in the removal of the racks. The rig design for the removal of the existing racks is specifically for the existing racks.
Lack of adequate inspection: The installer of the racks will develop a set ofinspection points which will eliminate any incidence of rework or erroneous installation.
Inadequate procedures: Southern Nuclear Operating Company will employ various operating procedures to address operations pertaining to the rerack effort, including, but not limited to, rack handling, upending, lifting, instalhtion, verticality, alignment, dummy gage testing, site safety, and ALARA compliance. Procedures for handling both the existing racks and new racks will be followed.
Table 3.5.1 provides a synopsis of the requirements delineated in NUREG-0612, and its intended compliance.
3.6 References
[3.5.1] " Spent Fuel Storage Module Corrosion Report", Brooks & Perkins Report 554, June 1,1977.
[3.5.2] " Suitability of Brooks & Perkins Spent Fuel Storage Module for Use in PWR Storage Pools", Brooks & Perkins Report 578, July 7,1978.
[3.5.3] "Boral Neutron Absorbing / Shielding Material - Product Performance Report", Brooks & Perkins Repon 624, July 20,1982.
[3.5.4] USNRC Letter to All Power Reactor Licensees, transmitting the "OT Position for Review and Acceptance of Spent Fuel Storage and llandling Applications", April 14, 1978, and later amended in January 18,1979.
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Table 3.3.1 BORAL EXPERIENCE LIST (Domestic and International)
Pressurized Water Reactors Vented Mfg.
Plant Utility Construction Year Bellefonte 1,2 Tennessee Valley Authority No 1981 Donald C. Cook Indiana & Michigan Electric No 1979/1992 Indian Point 3 NY Power Authority Yes 1987 Maine Yankee Maine Yankee Atomic Power Yes 1977/1994 Salem 1,2 Public Service Elec. & Gas No 1980/1994 Sequoyah 1,2 Tennessee Valley Authority No 1979/1992 Yankee Rowe Yankee Atomic Power Yes 1964/1983 Zion 1,2 Commonwealth Edison Co.
Yes 1980 Byron 1,2 Commonwealth Edison Co.
Yes 1988 Braidwood 1,2 Commonwealth Edison Co.
Yes 1988 Yankee Rowe Yankee Atomic Electric Yes 1988 Three Mile Is.1 GPU Nuclear Yes 1990 Connecticut Yankee Northeast Utilities Yes 1994 Fort Calhoun Omaha Public Power District Yes 1993 Beaver Valley 1 Duquesne Light Company Yes 1992 Shearon liarris (B)
Carolina Power & Light Yes 1991/1995 Boiling Water Reactors Browns Ferry 1,2,3 Tennessee Valley At 'lority Yes 1980 Brunswick 1,2 Carolina Power & Light Yes 1981 Clinton Illinois Power Yes 1981 Cooper Nebraska Public Power Yes 1979 Dresden 2,3 Commonwealth Edison Co.
Yes 1981 Duane Arnold lowa Elec. Light & Power No/Yes 1979/1993 i
J.A. FitzPatrick NY Power Authority No/Yes 1978/1988 E.1. Hatch 1,2 Georgia Power Yes 1981 Ilope Creek Public Service Elec. & Gas Yes 1985 Humboldt Bay Pacific Gas & Electric Yes 1986 Lacrosse Dairyland Power Yes 1976 Limerick 1,2 Philadelphia Electric No/Yes 1980/1994 l
Monticello Northern States Power Yes 1978 l
Peachbottom 2,3 Philadelphia Electric No 1980 l
Perry,1,2 Cleveland Elec. Illuminating No 1979 Pilgrim Boston Edison No/Yes 1978/1994 Susquehanna 1,2 Pennsylvania Power & Light No 1979 Vermont Yankee Vermont Yankee Atomic Power Yes 1978/1986 IIope Creek Public Service Elec. & Gas Yes 1989 LaSalle Unit 1 Commonwealth Edison Company Yes 1991 3-6
Table 3.3.1 (continued)
INTERNATIONAL INSTALLATIONS USING BORAL FRANCE 12 PWR Plants Electricity de France SOUTIl AFRICA Koeberg 1,2 ESCOM SWITZERLAND Beznau1,2 Nordostschweizerische Kraftwerke AG Gosgen Kernkraftwerk Gosgen-Daniken AG TAlWAN Chin-Shan 1,2 Taiwan Power Company Kuosheng 1,2 Taiwan Power Company MEXICO Laguna Verde Comision Federal de Electricidad Units 1 & 2 I
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f Table 3.3.2 1100 ALLOY ALUMINUM PHYSICAL PROPERTIES (TYPICAL)
Density 0.098 lb/cu. in.
2.713 gm/cc Melting Range 1190-1215 deg. F 643-657 deg. C Thermal Conductivity 128 BTU /hr/sq ft/deg. F/ft
_-(77 deg. F) 0.53 cal /sec/sq cm/deg. C/cm Coef. of Thermal 13.1 x' 10 in/in., *F Expansion 23.6 x 10 cm/cm, C (68-212 deg. F)
Specific heat 0.22 BTU /lb/deg. F
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(221 deg. F) 0.23 cal /gm/deg. C Modulus of 10x10 psi Elasticity Tensile Strength 13,000 psi annealed (75 deg. F) 18,000 psi as rolled Yield Strength 5,000 psi annealed (75 deg. F) 17,000 psi as rolled Elongation 35-45% annealed (75 deg. F) 9-20% as rolled Hardness (Brinell) 23 annealed 32 as rolled Annealing Temperature 650 deg. F 343 deg. C 3-8
Table 3.3.3 CHEMICAL COMPOSITION-ALUMINUM (1100 ALLOY)-TYPICAL 99.00% min.
Aluminum 1.00% max.
Silicone and Iron 0.05-0.20% max.
.05% max.
.10% max.
.15% max.
others each I
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- Table 3.3.4 BORON CARBIDE PROPERTIES (TYPICAL)
BORON CARBIDE CHEMICAL COMPOSITION, WEIGHT % -
Total boron 70.0 min.
B isotopic content in 19.45 min.
natural boron Total boron plus 97.0 min.
total carbon BORON CARBIDE PHYSICAL PROPERTIES BC Chemical formula 4
Boron content (weight) 78.28 %
Carbon content (weight) 21.72 %
Crystal Structure rhombohedral -
Density 2.51 gm/cc-0.0907 lb/cu. in.
Melting Point 2450 C 4442 F Boiling Point 3500 C-6332 F I
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1' Table 3.5.1 HEAVY LOAD HANDLING COMPLIANCE MATRIX (NUREG-0612)
CRITERION COMPLIANCE
- 1. Are safe 'oad paths defined for the movemt.1 of heavy loads to minimize the Yes potential ofimpact,if dropped on irradiated fuel?
- 2. Will procedures be developed to cover:
identification of required equipment, Yes inspection and acceptance criteria required before movement ofload, steps and proper sequence for handling the load, defining the safe load paths, and special precautions?
- 3. Will crane operators be trained and qualified?
Yes 4 Will special lifling devices meet the Not Applicable guidelines of ANSI 14.6-19787
- 5. Will non-custom lifting devices be installed and used in accordance with Yes ANSI B30.9-19717
- 6. Will the cranes be inspected and tested prior to use in rerack?
Yes
- 7. Does the crane meet the intent of ASME B30.2-1976 and CMMA-707 Yes 3-11
ENCLOSURE 2 RESPONSES TO RAI GROUP 1 QUESTIONS RADIOLOGICAL 1
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Question 1 Discuss how the increased number of fuel assemblies stored in the Vogtle Unit 1 SFP will affect the dose rates in adjacent accessible areas to the SFP (including any accessible areas below the SFP). You should describe how the does rates will differ both during I
storage and movement of spent fuel. State wiiether the increased storage capacity in the Vogtle Unit 1 SFP will necessitate any radiation zoning changes to any of the surrounding areas.
Response to Question 1 The smaller water gap between the new racks and the SFP wall (as compared to the original design) may result in increased doses in adjacent areas. Dose rates in areas adjacent to the SFP during transfer of spent fuel will not change because there will be no change in the manner in which fuel transfer occurs. It is not antic: pated that there will be any changes to the existing radiation zones even when the racks are full. The existing dose rate analyses for the Unit 2 rack and pool configuration bounds that for the new Unit I configuration.
Question 2 Provide a description of any sources of high radiation that may be in the Vogtle Unit 1 SFP during diving operations to remove the old SFP racks and install the new racks.
Discuss what precautions (such as use of TV monitoring, tethers, etc. ) will be used to ensure that the divers will maintain a safe distance from any high radiation sources in the SFP. Describe how you plan to monitor the doses received by the divers during the re-racking operation (e.g., use of dosimetry, alarming dosimeters, remote readout radiation detectors).
Response to Question 2 We will remove all known sources of11igh Radiation from the Unit-1 SFP and perform j
an extensive underwater radiation survey prior to allowing divers access to remove the old spent fuel storage racks. In addition the weir gate to the Cask loading pit that separates the Unit.1 & Unit-2 Pools and the weir gate to the Unit-1 transfer canal will be closed to isolate the Unit-1 SFP from all other liigh Radiation sources thereby ensuring that the divers maintain a safe distance from any liigh Radiation sources. We will employ the use of cameras, alarming tele-dosimeters, multibadge TLDs and voice comm'.inication to monitor the diver. In addition the diver will be fitted with a remote readout radiation monitoring instrument so that specific area surveys may be performed upon the request of IIcalth Physics. The diver will be fitted with a tether as a final safety measure.
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Question 3 Discuss what precautions you will take to ensure that the SFP racks purchased from
~ Maine Yankee have been adequately decontaminated to ensure that there are no hot spots or loose contamination that would result in unnecessary exposure to or contamination of the personnel unpacking and installing the SFP racks in the Vogtle Unit 1 SFP.
Response to Question 3 l
The area where the Maine Yankee racks will be unpacked will be set up to control any loose contamination that is present on the racks. Radiological surveys of the racks were sent along with the shipping paperwork and have been retrieved to use as a reference i
when unpacking the racks. Contamination controls will include the use of11 EPA ventilation, wrapping and decontamination where required. Ilot spots will be identified and verified through confirmatory surveys and workers will be briefed on methods to minimize their exposure. Continuous llealth Physics coverage will be provided throughout all aspets of thejob. All activities will take place in a controlled manner inside a Radiological Control Area.
Question 4 Your submittal does not include an analysis of the potential radiological consequences of a fuel handling accident in the fuel handling building, although such an analysis is contained in Chapter 15 of your FSAR. Verify that all of the assumptions used in the FSAR fuel handling accident analysis in the fuel handling building are still applicable.
Also verify that the resulting postulated thyroid and whole body doses at the Exclusion Area Boundary, Low Population Zone, and to the Control Room Operator as a result of a fuel handling accident are still valid.
Response to Question 4 l
The input parameters for a fuel handling accident in the fuel handling building currently shown in FSAR Tables 15.7.4-1,15.7.4-2,15.7.4-3,15A-2,15A-3, and 1SA-5 are unalTected by the change in SFP racks. Additionally, the method of analysis has not 1
changed; therefore, the radiological consequences shown in FSAR Table 15.7.4-4 remain J
applicable on replacement of the Unit I spent fuel racks.
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Question 5 Discuss how the storage of the additional spent fuel assemblies in the Vogtle Unit 1 SFP will affect the releases ofI-131 and tritium from the SFP.
Response to Question 5 The additional spent fuel assemblies stored in the Unit 1 SFP will increase the releases of I-131 and tritium from the spent fuel pool; however, FSAR Tables 12.2.2-1 and 12.2.2-2 note that only the tritium release is significant. The major source of tritium considered in the FSAR analyses is the accumulated inventory in the SFP water and connected systems (RCS, RWST, and RMWST) originating from operational tritium production in the RCS.
These sorirces are not affected by increased storage capacity in the SFP. Furthermore, the calculations of tritium releases include the assumption of boron recycle which has been discontinued, hence the analyses as currently reflected in the FSAR remain bounding.
Question 6 Discuss how the storage of the additional spent fuel assemblies will affect the releases of radioactive liquids from the plant.
Response to Question 6 No changes to the liquid radwaste system are required to support operation with the increased spent fuel storage capacity in Unit 1. As noted above in response to Question 5, no significant changes to the spent fuel pool releases is anticipated. Therefore, no increase in radioactive liquid releases due to the additional spent fuel storage capacity is anticipated.
Question 7 In Section 9.4 of your September 4,1997 request to add additional fuel storage racks to the Unit 1 SFP, you estimate a total estimated dose for the SFP re-racking operation of between 1.5 and 2.5 person-rem. Provide a dose breakdown byjob to show how you arrived at this dose estimate.
Response to Question 7 The original estimates were based on conversations with other plants who have performed re-racking operations. Based on a detailed review of the physical activities, durations, and the known/ expected dose rates associated with each activity we now expect to accrue approximately 4.3 person-rem. See Attachment 1 of Question 7 for details.
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Group 1 Question 7 Unit 1 Spent Fuel Pool Re-Rack Exposure Estimates Task Number Duration Person Dose Rate Total Worker (Hours)
-Ilrs.
(mrem /hr)
Dose s
(Person-mrem)
Preparation Work Assess / test gantry crane 4
200 800 0.25 200 Attach temp. crane 4
24 96 0.25 24 Load test W lift rig 4
24 96 0.25 24 Load test MY lift rig 4
24 96 0.25 24 Remove /decon old rig 4
48 192 2.0 384 Vacuum spent fuel pool 1
72 72 0.25 18 Re-Rack Build transfer pad new fuel pit 4
96 384 0.25 96 Build drainage pool 4
48 192 0.25 48 Build leveling pad 4
24 96 0.25 24 Unload replacement racks 4
312 1248 1.0 1248 Feet work 4
26 104 5
520
_ Dry drag testing 4
26 104 5
520 Leveling 4
26 104 2.5 260 i
Transfer to pool 4
13 52 0.5 26 Level rack 6
26 156 2.5 390 Drag test 3
52 156 0.25 39 QC acceptance 1
26 26 0.25 7
Equipment Removal and Clean-up Remove gantry crane 4
72 288 0.25 72 Remove transfer platform 4
48 192 0.25 48 Clean railroad bay 4
48 192 0.25 48 Health Physics coverage 2
650 1300 0.25 325 l
I TOTALS 5,946 4.345 l
hours person-Rem l
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ENCLOSURE 3 RESPONSES TO RAI GROUP 2 QUESTIONS CIVIL ENGINEERING AND GEOSCIENCE 1
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Question 1 With respect to the dynamic fluid-structure interaction analysis using the computer code, DYNARACK, in Reference 1, provide the following:
a) A composite stick model used in the dynamic analyses can not represent accurately and realistically the actual highly complicated nonlinear hydrodynamic fluid-rack structure interactions and behavior of the fuel assemblies and the box-type rack structures. Indicate whether or not you or your contractors actually have performed fluid-rack structure interaction analyses with 3-D FE plate, beam and fluid elements.
If you have, submit the results of the stick and continuum (plate / beam) model analyses. These results are needed to establish areas of maximum stresses, and then the detailed structural design.
b) Provide the results of any existing experimental study that verifies the correctness or adequacy of the simulation of the fluid couplin,, utilized in the numerical analyses for the fuel assemblies, racks and walls. If there is no full scale experimental study available, provide information to demonstrate that the current level of the DYNARACK code verification is adequate for engineering application.
c) Indicate whether you had any numerical convergency and/or stability problem (s) during the nonlinear, dynamic single-and multi-rack analyses using tne DYNARACK code. If there were any, how did you overcome the problem?
d) Indicate whether or not you have performed fluid-rack structure interactions where the fluid was treated as air to verify accuracy and capability of the DYNARACK code. If you have done this, provide the results of the study and any conclusions you have made from the study (i.e., numerical stability, convergence, etc.).
Response to Question 1 (a) and (b) 9 i
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l Response to Question 1(c)
No convergence problems were experienced during any of the simulations. Stability is achieved aner a certain minimum time step (for numerical integration) is established.
l The time step is chosen based on previous experience with the solver and is adjusted during the initial runs if solutions are not obtained. The numerical quadrature algorithm in DYNARACK requires very small time steps, which is computationally onerous, but has the desirable attribute of converging to the correct solution asymptotically as the time step is decreased.
Response to Question 1(d)
Over the years, Hohee has performed ad hoc (non Q.A.) comparisons of rack responses in the air and water media. When water is replaced with air, the fluid coupling terms vanish resulting in a decoupling of the motion of the modules from each other (unless they collide). Absence of water was found to lead to increased displacement and stresses.
Since there is no credible scenario wherein the fuel racks loaded with spent nuclear fuel in the Vogtle Unit 1 pool will be in the air medium, these racks have not been analyzed to establish their seismic response in air.
Question 2 Provide the time history input data in ASCII format for the operating basis earthquake (OBE) and design base earthquake (DBE) used in the rack analyses on a 3.5-inch diskette. Discuss how you developed the actual and target power spectra density (PSD).
Provide input data used for the target PSD development.
Response to Question 2 The requested diskette is provided which includes the time history input data for the OBE and DBE events at the elevation ofinterest.
l The time histories utilized in evaluating the fuel racks for the Vogtle Unit 1 pool are identical to those used in qualifying the racks for the Vogtle Unit 2 (west) pool during the 1987-89 rerack project. Both pool structures (Unit I and Unit 2) are architecturally identical and have the same floor response spectra. Information on development of the actual and target PSD was retrieved from VEGP response to a similar question from the Staff during the licensing of Unit 2. Attachment 1 of Question 2 contains the response submitted by Georgia Power Company letter (File: X7BC35, Log: GN-1475) dated July 21,1988, under Docket 50-425, Construction Permit Number CPPR-109).
E3-6
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Question 3 Run I.D. dvog-pi.sxt in Table 6.5.86 of Reference I shows horizontal displacements of 4.8 inches and 0.3 inch for the top and bottom of a rack structure, respectively. Provide a complete deformation shape of the rack from the bottom to the top of the rack.
Response to Question 3 An examination of the kinematic data from the fuel rack analysis carried out in support of this licensing application indicates that the motion of the rack is essentially that of a rigid structure. In other words, the clastic component of the displacement is quite small.
The large horizontal displacement at the top of the rack is mainly due to the tilting motion of the rack. The displacement profile of the rack from the top to the bottom is nearly l
linear. Only a small fraction of the total displacement is attributed to bending and torsion. Sirailar behavior has been observed in other rack dynamic simulations. For example, an analysis presented in Docket No. 50-390 for par racks, which were installed at Watts Bar nuclear plant, showed that only 2% of the total displacement was from the elastic effect. The racks at Vogtle Unit I are also designed by par, and their construction is nearly identical.
Question 4 The analysis in Reference 1 does not show any vertical displacement (rack uplift).
Indicate whether there is a vertical displacement of the rack in the analysis. If there is, what is the maximum impact load between rack and steel liner? Explain how the DYNARACK code handles a separation (uplift) between a rack and liner. Provide any comparison study between DYNARACK analysis and experimental test that confirms the l
DYNARACK's capability to represent an uplift (separation) in an analysis.
Response to Question 4 Please see the response to question 5 for vertical displacements. Maximum impact load between rack and steel liner is listed as maximum total vertical pede,tal load in table 6.5.94, reference 1.
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A lucid description of the methodology to treat interface friction and separation is provided in The Component Element Method in Dynamics, by Levy and Wilkinson, McGraw-Hill Inc. (1976). Unfortunately, no Q.A. validated experimental data on free standing rack motion is available for comparison with DYNARACK results Question 5 A brief summary of the whole pool multiple rack (WPMR) analysis is shown in Table 6.5.94 of Reference 1. The xesults show that the maximum rack-to-rack impact force, rack-to-wall impact force and horizontal displacement are approximately 167,500 lbs., O lb. (no impact) and 5.19 inches, respectively. Provide detailed information in a tabular form for the racks J, S, V, M, O, N, D, G and Y.
a) The maximum kinetic energy between the racks during the impact including masses, spectral velocities, natural frequencies and impact forces.
b) The maximum displacements in both hoix ital and the vertical directions.
c) The physical dimensions of the racks, the gaps between the racks and the gaps between the racks and the walls.
Response to Question 5(a)
The kinematic energy of the system is not explicitly calculated by the computer code.
The velocities, displacements, and accelerations of all generalized coordinates (degrees-of-freedom) are calculated internally in the code, but not saved. A special ; urpose post-processor will have to be written, debugged, and Q.A. validated to extract it. formation of the component kinematic energy, velocities, accelerations, etc.. Since the analysis method does not require decomposition of the element motion into discrete harmonics, frequencies and spectral velocities are not germane to the DYNARACK solution. The rack-to-rack impact forces, however, are computed and saved. The following table lists the maximum impact forces for the racks that are identified in Question 5.
E34
MAXIMUM RACK TO RACKIMPACTFORCES RUN NO.
IMPACT LOCATION IMPACT FORCE,
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Rack M - Rack O (top) 46880 1
Rack O - Rack H (top) 141000 1
Rack N-Rack I (top) 51340 1
Rack N - Rack D ( < p) 64590 1
Rack D - Rack T (top) 15670 1
Rack D-Rack G (top) 103500 1
Rack G - Rack Y (top) 95380 1
Rack V - Rack H (top) 104400 1
Rack S - Rack L (top) 29220 2
Rack M - Rack V (top) 59650 3
Rack G - Rack Y (top) 67330 Note that rack J does not impact any of the surrounding racks.
Response to Question 5(b)
The following tables contain the maximum horizontal and vertical displacements of every rack in the pool.
l 1
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RUN NO.1 (SSE, g = 0.8)
RACK NO.
MAXIMUM MAXIMUM IiORIZONTAL (per WPMR)
VERTICAL DISPLACEMENT (inch)
DISPLACEMENT (inch) l M
1 0.423 2.047 0
2 0.542 2.212 N
3 0.339 1.515 D
4 0.548 2.253 G
5 0.297 2.269 Y
6 0.430 1.874 F
7 0.467 2.263
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0.309 1.382 11 9
0.596 2.639 I
10 0.442 1.806 T
11 0.241 1.803 E
12 0.414 1.487
- U 13 0.505 1.820 B
14 0.623 2.247 S
15 0.502 2.283 L
16 0.635 2.538 P
17 0.624 3.228 C
18 0.425 3.115 K
19 0.27i 1.967 X
20 0.391 1.958 R
21 0.550 1.988 J
22 0.429 1.556 A
23 0.354 2.192 W
24 0.330 1.588 Q
25 0.391 2.885 Z
26 0.323 2.133 i
I E3-10
RUN NO. 2 (SSE, p = 0.2)
RACK NO.
MAXIMUM MAXIMUM HORIZONTAL (per WPMR)
VERTICAL DISPLACEMENT (inch)
DISPLACEMENT (inch)
M 1
0.235 2.340 0
2 0.210 1.833 N
3 0.191 1.453 D
4 0.133 1.599 G
5 0.131 1.496 Y.
6 0.353 1.242 F
7 0.315 2.648 V
8 0.328 1.902 11 9
0.277 1.841 I
10 0.315 2.099 T
I1 0.214 2.103 E
12 0.301 2.535 U
13 0.379 3.384 B
14 0.552 2.587 S
15 0.301 2.613 L
16 0.184 1.037 P
17 0.215 1.465 C
18 0.259 1.623 K
19 0.222 1.607 X
20 0.300 1.514 R
21 0.311 2.306 i
J 22 0.469 2.556 A
23 0.257 1.471 l
-W 24 0.330 1.779 Q
25 0.272 2.024 Z
26 0.185 2.053 l-E3-11
RUN NO. 3 (SSE, g = Gaussian Dist.)
RACK NO.
MAXIMUM MAXIMUM HORIZONTAL (per WPMR)
VERTICAL DISPLACEMENT (inch)
DISPLACEMENT (inch)
M 1
0.357 1.644 O
2 0.431 2.006 N
3 0.342 1.466 D
1 0.377 1.528 G
5 0.366 1.969 Y
f 0.374 1.909 F
-7 0.523 2.092 V
8 0.378 2.216 11 9
0.480
!.898 I
10 0.479 1.814 T
11 0.242 1.456
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13 0.411 1.762 B
14 0.568 2.713 S
15 0.388 1.695 I
1, 16 0.349 1.563 P
17 0.444 1.648 C
18 0.330 1.822 K
19 0.461 2.512 X
20 0.407 1.766 R
21 0.394 1.906 J
22 0.288 1.278 A
23 0.273 1.517 W
24 0.323 1.390 Q
25 0.805 5.193 Z
26 0.684 4.384 ~
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I E3-12
______________________________a
RUN NO. 4 (OBE, n = Gaussian.Dist.)
RACK NO.
MAXIMUM MAXIMUM IIORIZONTAL (per WPMR)
VERTICAL DISPLACEMENT (inch)
DISPLACEMENT (inch)
M 1
0.174 0.889 0
2 0.372 1.799 N
3 0.438 2.067 D
4 0.379 1.743 G
5 0.371 2.105 Y
6 0.819
?.985 F
7 0.413 3.001 V
8 0.308 1.356 11 9
0.308 1.379 I
10 0.535 2.105 T
11 0.209 1.168 E
12 0.303 1.256 U
13 0.418 1.740 B
14 0.454 1.911 S
15 0.323 1.739 L
16
.0.351 1.783 P
17 0.435 1.676 C
1 0.407 1.839 K
19 0.303 1.701 X
20 0.358 2.026 R
21 0.497 2.184 J
22 0.439 1.979 A
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23
'0.373 1.527 W
24 0.652 2.466 Q
25 0.417 2.242 Z
26 0.466 2.547 Note that while application of the OBE shows that the maximum rack response in the pool occurs for the SSE case, it does not preclude that some individual racks may experience larger maximum displacements for the OBE case. This is due to the complex E3-13
interaction between the fluid coupling forces, the frictional forces, and the rack impact forces. The absolute maximum excursion does occur for the SSE event.
Response to Question 5(c)
The figure in Attachment 2 of Question 1 shows the physical dimensions of the racks and the gaps. The overall height of the storage racks is 176.75 inches. The rack pedestals, which are 5.125 inches high, rest on 2 inch thick stainless steel bearing pads.
Question 6 With respect to the analytical simulation of the rattling fuel assembly impacting against the cell for the racks J, A, M, V, N, Q and Z, provide the following:
l a) llow did you calculate the magnitude of the largest impact force and the location of the impact in the fuel assembly and the cell wall?
b) 110w did you determine and analyze the fuel assembly and cel vall integrity?
c) Discuss the considerations given to the effects of the fluid between the fuel assembly and cell wall during the interactions.
d) Available experimental studies that verify the reasonableness of the numerical simulation adopted to represent the fuel assembly and cell wall interaction.
e) The maximum kinetic energy between the fuel assembly and cell wall during the impact including masses, spectral velocities, natural frequencies and impact forces.
f) The maximum displacements in both horizontal and vertical directions.
g) The physical dimensions of the fuel assemblies, and gaps between the fuel assemblies and the cell walls.
Response to Question 6 e
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DYNARAC produces a complete time history of the loadings within these non-linear compression only gap / spring elements and archives the results for later review or post-processing. Post-processors enable scanning of the large number of time steps for instants where loads are actually present and allows for direct retrieval of the bounding values.
b)
The integrity of the cell wall is evaluated by analyzing the local stresses developed in the wall panel under the lateral dynamic load from rattling of the assemblies. Even though subsection NF of the ASME Code does not place any limits on the local stresses in Class 3 structures, the acceptance criterion for local cell wall stresses is set equal to the yield strength of the cell wall material. The maximum impact load on the cell wall from any simulation does not cause any yielding of the cell wall material.
The fuel assembly consists of an assemblage of fuel rods, which are smail diameter hollow tubes. By virtue of the fuel rod cylindrical geometry, a fuel rod j
is an order of magnitude more lateral load resistant than the cell wall, which is a thin wall " plate" stock.
The integrity assessment of the cell wall, therefore, is traditionally accepted as sufficient to qualify the fuel assembly against rod failure as well. Accordingly, a damage evaluation of the fuel rods was not performed.
c)
The 11uid coupling between the fuel assembly and the cell wall is treated by inertial coupling in the system kinetic energy. The methodology is adapted from the classical fluid mechanics treatment by Fritz in "The Effects of Liquids on the Dynamic Motions ofimmersed Solids" A copy of this article is provided as of Question 6 along with additional discussion provided by Singh and Soler in " Dynamic Coupling in a Closely Spaced Two-Body System Vibrating in Liquid Medium: The Case of Fuel Racks," provided as Attachment 2 of Question 6.
d)
There are no experimental studies auilable that correlate the simulated behavior of the fluid coupling between assembhes and the cell wall during dynamic events with actual data. Ilowever, since the methodology for modeling this coupling is taken from classical fluid mechanics, it is widely accepted.
e)
The kinematic energy of the system is not explicitly calculated by the computer code. The velocities, displacements, and accelerations of all generalized E3-15
coordinates (degrees-of-freedom) are calculated internally in the code, but not saved. A special purpose post-processor will have to be written, debugged, and Q.A. validated to extract information of the component kinematic energy, velocities, accelerations, etc.. Since the analysis method does not require decomposition of the element motion into discrete harmonics, frequencies and spectral velocities are not germane to the DYNARACK solution. The fuel-to-cell wall impact forces, however, are computed and saved. The following tables report the maximum values for every rack in the pool.
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RUN NO.1 (SSE, g = 0.8)
RACK NO.
MAXIMUM CELL TO FUEL (per WPMR)
IMPACT FORCE PER CELL (pounds)
M 1
3138 O
2 2958 N
3 3041 D
4 2703 G
5 3314 Y
6 3317 F
7 3083 V
8 3902 H
9 3000 I
10 3194 T
11 3037 E
12 3240 U
13 3095 B
14 3055 S
15 3303 L
16 3214 P
17 3392 C
18 3229 K
19 2770 X
20 2812 R
21 3916 J
22 3232
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A 23 3142 W
24 3089 Q
25 3472 Z
26 3527 l
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RUN NO. 2 (SSE, n = 0.2)
RACK NO.
MAXIMUM CELL TO FUEL (per WPMR)
IMPACT FORCE PER CELL (pounds)
M 1
3194 0
2 2722 N
3 3027 D
4 2722 G
5 3203 Y
6 3126 F
7 2875 V
8 3069
_ H 9
2569 I
10 3222 T
11 2481 E
12 2518 U
13 2841 B
14 2981 S
15 2803 L
16 3250 P
17 3321 C
18 3312 K
19 2520 X
20 2333 R
21 3104 3
22 3214 A
23 2892 W
24 2696 l
Q 25 2888 Z
26 3388 E3-18
l 1
l l
RUN NO. 3 (SSE, y = Gaussian Dist.)
RACK NO.
MAXIMUM CELL TO FUEL (per WPMR)
IMPACT FORCE PER CELL (pounds) i M
i 3111 0
2 2944 N
3 4000 D
4 2981 G
5 2759 Y
6 3412 F
7 3270 V
8 3902 H
9 3402-1 10 3277 T
11 3425 E
12 2870 U
13 3111 B
14 3370 S
15 3250 L
16 3178 P
17 3250 C
18 3125 K
19 3208 X
20 2833 R
21 3958 J
22 4000 A
23 2607 W
24 3607 Q
25 3416 Z
26 3527 E3-19
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L l
RUN NO. 4 (OBE, n = Gaussian Dist.)
RACK NO.
MAXIMUM CELL TO FUEL (per WPMR)
IMPACT FORCE PER CELL (pounds)
M 1
2486 0
2 2319 N
3 2444 D
4 2481 G
5 2500 Y
6 2507 F
7 2687 V
8 2263 11 9
2722 1
10 2666 T
11 2555 E
12 2740 U
13 2841 B
14 2611 S
15 2714 L
16 2517 P
17 2464 C
18 2645
^
K 19 2812 X
20 2958 R
21 2354 J
22 2803 A-23 2892 W
24 2785 Q
25 2833 Z
26 2619 E3-20
I f)
The motion of the stored fuel is constrained by the fuel rack cell walls. Therefore, the maximum horizontal fuel assembly displacements are bounded by the maximum fuel rack displacements. The maximum vertical displacement is equal to the maximum rack vertical displacement because the degrees-of-freedom are coupled in the vertical direction. This assumption is consistent with previous rack dynamic simulations. Please see the response to question 5(b) for the maximum fuel rack displacements, g)
The physical dimensions of the fuel assembly are approximately 8.424 in x 8.424 in x 160 in. The inner dimension of a storage cell is 8.75 in (square). The mean gap dimension between the fuel assembly and the cell wall is 0.163 in.
Question 7 The staffis also reviewing an application for the modification of the spent fuel pool storage racks for Waterford 3 submitted by Entergy Operations, Inc. (Reference 2). In Reference 2, your contractor, llOLTEC, predicts the maximum displacement of 1.13 inches for a rack in the spent fuel pool structure at Waterford 3, which is much small than your maximum displacement of 5.19 inches for a racks at Vogtle. The difference between these two displacements is significant although the staff understands that both racks were designed under different structural considerations. Identify the factors that cause such a large difference in displacements and provide detailed technical discussions on how those factors are contributing such a large difference. Provide the technical explanations that the difference is not due to the lack of the DYNARACK's ability to model a nonlinear free standing fluid-structural system.
Response to Question 7 Differences in response between spent fuel racks in Vogtle and Waterford nuclear plants can be ascribed to one or more of the following effects:
1.
Differences in seismic input 2.
Differences in rack pedestal footprint 3.
Dif*erences in spent fuel pool fluid gaps in general rack response at a given site increases monotonically with increasing seismic intensity, with decreasing rack pedestal footprint, and with increasing fluid gaps between racks and between racks and pool walls.
On the basis of the above remarks, we can evaluate differences between responses (i.e.,
maximum rack displacements) at the two plants in question. The Waterford data used in the comparion were provided by 11oltec who performed the Waterford analyses.
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)
I With regard to seismic input both plants have horizontal SSE floor spectra with zero period acceleration (ZPA) approximately equal to 0.4g. A comparison of power spectral densities reveals that the Vogtle target floor spectral response extends over a broader frequency range than does the corresponding Waterford power spectral density function.
In addition, the peak power spectral density is approximately ten times greater in the Vogtle plant (in units ofin /sec'). Therefore, we can qualitatively assert that the input 2
seismic loading driving the spent fuel rack response is " stronger" at Vogtle than the corresponding input is at Waterford.
i With regard to pedestal envelopes, we note that for rack Q (Vogtle), the rack envelope is 62 in x 62 in. For rack B5 (Waterford), the rack envelope is 89.25 in x 95.95 in. The pedestals on both racks are located under the corner cells. Therefore, we expect that the l
Vogtle rack will have less resistance to rotational motion because of the smaller
" footprint" of the pedestals.
Finally, the table below lists the rack-to-rack gaps and the rack-to-wall gaps for the two plants.
Vogtle Unit 1 Waterford Unit 3 Rack-to-Rack Gap 2 in 0.75 in Rack-to-Wall Gap (min.)
19 in 1.25 in Rack-to-Wall Gap (max.)
53 in 2.25 in it is clear that the fluid coupling effect will be smaller in the Vogtle plant and lead to larger rack motions.
We conclude on the basis of the above comparisons that it is expected that the motion of spent fuel racks in the Vogtle spent fuel pool will bound the motions in the Waterford spent fuel pool. Thus, the maximum rack movements of 5.19 inches at Vogtle Unit I and 1.13 inches at Waterford Unit 3 predicted by IIoltec is consistent with expectations and provides additional confirmation that the DYNARACK code with its fluid coupling model, predicts results consistent with physical reality.
Question 8 Your analysis results show that there were rack-to-rack impacts. Indicate whether you are planning to install a support system to minimize displacement and impact force between the rack-to-rack.
Response to Question 8 The high-density racks for Vogtle Unit 1 are designed for installation as freestanding structures. Numerous plants have installed racks in freestanding mode where the licensing basis analyses had indicated inter-rack and/or rack-to-wall impacts. Examples from the E3-22
past are: Byron 1 and 2 (1988, Docket Nos. 454 and 455); Braidwood 1 and 2 (1988, Docket Nos. 456 and 457); St. Lucie Unit 1 (1987, Docket No. 50-335); Diablo Canyon 1 and 2 (1987, Docket Nos. 50-275 and 323); and Ft. Calhoun (1994, Docket No. 50-285).
Impact during seismic events is a natural corollarf of a freestanding structure. At minimum, during seismic events, the fuel assemblies rattle inside the storage cavity and rack pedestals' compression forces change with time. Pedestal lift-off and impact are also more of a rule than an exception. Rack-to-rack impact is also observed in a significant number of cases. None of these impact forces would lead to an adverse effect on safety if their magnitudes are conservatively quantified and if their consequences to the rack structure are carefully examined. The Vogtle Unit I racks have been subjected to an exhaustive set of dynamic and stress analyses to ensure that the safety conclusions stated in Reference 1 are accurate. Where rack-to-rack impacts occur, it has been established that there is no effect on the structure in the region of active fuel, which insures that the safety function of the rack will not be impaired. Analyses show that the region oflocal plasticity due to inter-rack impact is far removed from the active fuel region in the Vogtle Unit I racks.
The magnitudes of the impact forces suggest that there is no need to add any fixity to the racks in the fuel pool.
Question 9 Figures 6.5.31 through 6.5.34 of Reference I show negative hydrodynamic pressures.
Explain the physical meaning of the negative pressures.
Response to Question 9 The pressure values shown in Figures 6.5.31 through 6.5.34 of Reference 1 do not represent absolute pressures. The plots indicate the change in pressure relative to the hydrostatic load. The pressure values in Figures 6.5.31 through 6.5.34 must be added with the hydrostatic load to obtain the total pressure (static plus dynamic) in the fluid channel. A negative hydrodynamic pressure indicates that the total pressure is less than the coincident hydrostatic pressure.
, uestion 10 Q
Figures 6.5.32 through 6.5.34 of Reference I show discontinuity of hydrodynamic pressures. Explain the physical meaning of the discontinuity.
Response to Question 10 The hydrodynamic pressure is not fully continuous along the length of the spent fuel pool wall because the spent fuel racks are discrete objects separated by fluid gaps. Adjacent racks can have different velocities and accelerations, which result in different local E3-23
T c
pressures on the pool wall. The algorithm used to compute the fluid pressure does not compute the actual sudden changes across the gaps between racks; it computes the values up to each end of a rack along the channel.
l Question 11 Submit your calculations for the accident analysis mentioned in Chapter 7 of the submittal (Reference) including detailed information (i.e., input loading parameters, material properties and characteristics, impact velocity, mass, force, boundary conditions, engineering assumptions made, etc.).
Response to Question 11 l
The calculations for the accident analysis are provided as Attachments 1 and 2 of Question 11 (proprietary).
Question 12 Southern Nuclear Operating Company (SNOC) is planning to install the storage racks that have been used by the Maine Yankee Atomic Power Company (MYAPC). Discuss the quality assurance and inspection programs to confirm the structural integrity of the racks (i.e., material strength and property changes, corrosion and material creep / hardening due to thermal loadings and radiations, etc.). Submit a summary report of the structural degradations that deviate from the licensed structural conditions of the MYAPC, if there are any.
Response to Question 12 The lifling, packaging, and shipping of the MYAPC racks conformed to the original rack vendor guidelines. At VEGP, the racks will be unpacked and upended using approved procedures and devices specifically designed for handling racks. A Quality Control (QC) receipt inspection of each rack will be performed. The inspection will include visual examinations and pre-installation drag testing of 20% of the storage cells. The racks will be installed in accordance with approved procedures and with the use of devices specifically designed for handling racks. Post-installation drag testing will be performed on all storage cell locations. QC inspections will be an integral part of the above-described process at VEGP. Any structural degradation found during Q.C. receipt inspections will be dispositioned in accordance with approved procedures.
l E3-24
Question 13 With respect to the spent fuel pool (SFP) structural analysis using the ANSYS computer code presented in Chapter 8 of Reference 1:
a)
Provide a plan view of the SFP and physical dimensions of the reinforced concrete slab and walls, liner plate and liner anchorage.
b)
Provide a copy of the mesh used in the analysis.
c)
Describe the boundary conditions used, and indicate them in the mesh.
d)
Provide the material properties (i.e., modulus of elasticity, shear modulus, Poisson's ratio, yield stress and strain, ultimate stress and strain, etc.) of the liner plate and liner anchorage at the temperature of 195 F used in the analysis, e)
Describe the applied loading conditions including the magnitudes, and indicate their locations in the mesh.
f)
Justifications of using the COMBIN14 element to model the soil-structure interactions, g)
In Chapter 6 of Reference 1, you emphasized the importance of using the nonlinear, time history analysis. Ilowever, in Chapter 8, you used equivalent static analysis. This is contradictory. Provide justifications for using the equivalent static analysis. What were the magnitudes of the equivalent static accelerations on all masses you used in the analysis?
h)
Explain how the interface between the liner and concrete slab is modeled, and also, how the liner anchors are modeled. Provide the basis for using such modeling with respect to how they accurately represent the real structural behavior.
i)
Provide in a tabular form the complete analysis results including out-of-plane and shear stress, strain, deflection and reaction forces for the liner and liner plate and liner anchor under the maximum temperature loading condition (195 F) during an accident. Also, provide the analysis results that demonstrate that there is no local buckling on the liner plate c ner plate separation from the liner anchor (i.e., bifurcation point, mode shape,
- .).
1 j)
Provide all your input to and the results of the SFP analysis including the thermal and cracked section analysis.
E3-25
I Response to Question 13(a)
Attached are plan (AX1D09A04) and elevation (AX1D09A05) drawings of the Spent Fuel Pool, Attachments 1 and 2, respectively, of Question 13.
The reinforced concrete slab is 6 feet thick. The north fuel pool wall is 6 feet 6 inches thick. The south and east fuel pool walls are 6 feet thick. The west fuel pool wall is 5 feet thick. The top of the spent fuel pool walls is at the refueling floor, elevation 220 feet.
The top of the spent fuel pool floor slab is at elevation 179 feet.
The liner consists of 10 feet x 8 feet x % inch thick plates. The anchorage consists of rows of % inch x 2 inch backing strip bars at each edge and in the center of the plates, running parallel to the 8 foot side of the plates. These backing strip bars are anchored into the concrete with % inch diameter x 4 inch long headed studs spaced I foot on centers. The plates are continuously welded to the backing strip at the seams and also plug welded at each stud location on the backing strip rows located at the center of the 10 foot span of plate.
Response to Question 13(b)
Attached are two overall isometric views (A2 and A3 of 34), Attachments 3 and 4 respectively of Question 13, one looking South East and the other looking North West, of the finite element model for the Spent Fuel Pool Structure with the mesh indicated. To provide additional detail, plots of the fuel pool floor and the north, south, east, and west fuel pool walls with the mesh indicated are also provided (A8, A9, A10, Al1, and A12 of
- 34) in Attachments 5 through 9 of Question 13.
Response to Question 13(c)
Transnational linear springs are used to simulate the soil conditions that exist below the basemat. These springs are attached to each node at the pool floor elevation west of the electrical tunnel (Room R-C06, Section A on Attachment 2 of Question 13), to each node at the floor of the electrical tunnel, and to each node at the electrical tunnel floor elevation east of the electrical tunnel.
Along the axis of building symmetry, the centerline of the two-unit plant, symmetrical boundary conditions were used for vertical and north-south loads, and anti-symmetrical boundary conditions were used for east-west loads. Symmetrical boundary conditions include imposing zero translation in the east-west direction and zero rotation about the global vertical and north-south axes. Anti symmetrical boundary conditions include imposing zero translation in the north-south and vertical directions, and zero rotation about the global cast-west axis. These boundary conditions are applied to each node in the western most plane of the model.
E3-26
I I
Response to Question 13(d)
The material properties of the liner plate and liner anchorage are as follows:
I e Material:
Stainless Steel, Type 304L l
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Modulus of Elasticity:
27.6 x 10 psi i
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0.3 6
Shear Modulus:
10.6 x 10 psi Yield Stress:
21,300 psi e
e Yield Strain:
7.72 x 10" Ultimate Stress:
66,200 psi e Ultimate Strain:
0.38 Response to Question 13(e)
The applied loadings are dead, live, seismic, and thermal.
Dead load The structure gravity load was applied as the mass of the finite elements multiplied by the acceleration due to gravity.
The weights of the fuel handling building superstructure were applied as nodal loads along the top of the north and south exterior walls and at column locations on the east floor.
3 2
The hydrostatic pressure load equal to 0.0624 kips /ft x 39.5 feet of water = 2.46 kips /ft was applied to the pool floor and varied linearly on the pool walls from a maximum at the floor to zero at the top of the water level.
2 The static lateral soil pressure varying linearly from a value of 7.44 kips /ft at the bottom 2
to a value of 5.60 kips /ft at the top was applied to the electrical tunnel wall that is in contact with the soil.
Major equipment weights including the new racks at 2,539 kips were applied as pressure loadings to the pool floor. Fuel cask weight at 136 kips was applied as pressure loading to the refueling floor. Exhaust and filter units at 60 kips were applied as pressure loading E3-27
to the refueling floor. A heat exchanger at 52 kips was applied as a pressure loading to floor elevation 200 feet. New fuel racks at 324 kips were applied as a pressure loading to floor elevation 205 feet.
2 A uniform dead load equal to 50 lbs/ft was applied to the non-submerged floor area throughout the building.
Live load 2
A uniform live load equal to 100 lbs/ft was applied to the non-submerged floor area throughout the building.
The live load from floors in the fuel handling building superstructure were applied as nodal loads along the top of the north and south exterior walls and at column locations on the east floor.
Seismic load Equivalent static acceleration factors equal to the zero period accelerations from the floor response spectra for elevation 220 feet (refueling floor),0.25g for OBE east-west,0.20g for OBE north-south,0.29g for OBE vertical,0.39g for SSE east-west,0.30g for SSE north-south, and 0.43g for SSE vertical were applied to all dead load components including 25% of the live load, except for lateral soil pressure and pool water (hydrodynamic loads for pool water and dynamic lateral soil loads are described below.)
2 The dynamic lateral soil pressure linearly varying, for OBE from a value of 0.76 kips /ft 2
at the bottom to a value of 0.92 kips /ft at the top, and for SSE from a value of 1.19 2
2 kips /fl at the bottom to a value of 1.44 kips /ft at the top, was applied to the electrical tunnel wall that is in contact with the soil.
The peak time history result from the whole pool multi-rack seismic analysis was applied as an equivalent static load distributed equally among nodes on the fuel pool floor. See Figures 6.5.25 - 28 and 6.5.36 - 44 of the Licensing Modification Report for magnitudes.
Hydrodynamic loads based on TID-7024 " Nuclear Reactors and Earthquakes" were applied as pressures on the pool floor and walls. For OBE north-south magnitudes vary from 0.702 psi at the top of the water level to 3.715 psi at the bottom of the pool. For OBE east-west magnitudes vary from 0.965 psi at the top of the water level to 5.521 at the bottom of the pool. For SSE north-south magnitudes vary from 1.053 psi at the top of the water level to 5.573 psi at the bottom of the pool. For SSE east-west magnitudes vary i
from 1.505 psi at the top of the water level to 8.613 psi at the bottom of the pool.
Additional hydrodynamic loads based on the peak response from whole pool multi-rack seismic analysis were applied as equivalent static pressures on the pool walls. See Figures 6.5.31 - 34 of the Licensing Modification Report for magnitudes.
E3-28
l i
Thermal Loading l
l For the purpose of the structural analysis, the normal operating condition consists of the mean bulk concrete temperature of 115.5 F and the concrete temperature differential of 111 F applied to pool walls and floor.
For the purpose of the structural analysis, the accident condition consists of the mean bulk concrete temperature of 127.5 F and the concrete temperature differential of 135 F applied to pool walls and floor.
Response to Question 13(f)
The ANSYS COMBIN14 element was used in the spent fuel pool finite element model as a transnational linear " soil spring" to represent the soil response during various loading conditions. The value of the " soil spring" stiffness is calculated on the basis of the Winkler foundation model, that is K = Nodal Contributory Area x k,, where k,is the coefficient of subgrade reaction.
The Winkler foundation model is the most common procedure used in modeling the elastic behavior of the foundation soil. In ACI Committee 336 publication ACI 336.2R-88," Suggested Analysis and Design Procedures for Combined Footings and Mats," the committee states "the Winkler foundation model can be improved, but the increased computational complexity is not warranted when the soil properties are taken into account." Therefore, the use of the Winkler foundation model to evaluate the soil-structure interaction is deemed justified for this evaluation.
Response to Question 13(g)
Chapter 6 of the Licensing Modification Report deals specifically with the structural analysis of the racks. Nonlinear time history analysis is important to accurately predict the complex behavior of the freestanding racks and their interaction with the water in the pool.
Chapter 8 deals specifically with the analysis of the spent fuel pool structure. The spent fuel pool structure is designed to behave linearly. Equivalent static analysis using zero period accelerations of the floor response spectra for elevation 220 feet (refueling floor) is conservative. The OBE static acceleration factors are 0.25g east-west,0.20g north-south, and 0.29g vertical. The SSE static acceleration factors are 0.39g east-west,0.30g north-
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south, and 0.43g vertical.
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j E3-29
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Response to Question 13(h)
The liner in the Vogtle Unit 1 pool is assembled from austenitic steel plate which is seam welded along the contiguous edges of the plates resulting in a sealed container geometry to hold pool water. The seam weld lines are also locations of anchor. The integrity analysis of the pool liner consisted of the following evaluations:
1 J
The analyses show that the maximum liner in-plane stress is well below the ultimate limit for the liner material, and the cumulative damage factor (total fatigue expenditure under 6 SSE events)is less than 8% of the allowable value per the ASME Code. The stresses at the anchor locations are much lower because the effect of the locally applied load attenuates away from the loaded patch.
Response to Question 13(i)
The scenario of buckling under a limiting thermal condition in the pool postulated by the staff, is indeed likely. Ilowever, cracking of the liner under one such event is not possible because thermal stress is a self-limiting stress, formally referred to as " secondary stress"in the ASME Code. Secondary stresses, by definition, can not cause actual metal failure unless the plastic strain exceeds the endurance limit for Type 304 austenitic stainless steel material; the ultimate strain is 0.38. Such a large strain is considered to be unachievable in the Vogtle Unit 1 pool because of the temperatures involved (195 F max.)
Because of the above reasons, a finite element analysis of the liner to investigate thermal buckling induced failure was not carried out in support of the license application.
E3-30
Response to Question 13(i)
Input to the spent fuel pool analysis is the fuel handling building geometry, the boundary conditions as described in the answer to Question 13(c) above, and the loadings as described in the answer to Question 13(e) above.
Results of the spent fuel pool analysis are shown in Atiachment 10 of Question 13.
Question 14 Provide the locations of the leak chase system with respect to the locations of the racks and pedestals.
Response to Question 14 The leak chase system location in the Unit 1 SFP was considered in the placement position of the racks in the pool. The final placement is shown in Figure 2.1.1, Vogtle Pool Layout, of the Modification Report. The dashed line in this figure indicate the pool leak chase. The racks were placed in the pool in a configuration to avoid placing bearing pads on or too close to a leak chase. The leak chase positions were accounted for in the bearing pad analysis discussed in Section 8.6 of the Modification Report.
Question 15 Describe the method ofleak detection in the fuel pool structure. Ilow are leaks monitored? Is there any existing leakage?
Response to Question 15 The leak detection system for the SFP is described in FSAR section 9.3.3.2.3.5, Liner Plate Leakage Detection. Valved lines connected behind the spent fuel pool liner plate discharge to the leak detection pit in the fuel handling building. A dripping 3/4-in.
line would indicate the existence of a leak. The pits are monitored daily and currently there is no leakage.
Question 16 Discuss the quality assurance and inspection programs to preclude installation of any l
irregular or distorted rack structure, and to confirm the actual fuel rack gap configurations with respect to the gaps assumed in the DYNARACK analyses after installation of the racks.
(
E3-31
Response to Question 16 Procedures are implemented upon receipt of a new rack on site to ensure that the rack is a quality product meeting the requirements of the USNRC OT Position Paper. The rack is first inspected for any obvious damage potentially caused by earlier handling or shipping.
The rack is inspected for any scratches, dents, or signs of environmental exposure induced degradation. Once the rack is offloaded and upended, another procedure is implemented for pre-installation dry testing. A percentage of cells are drag tested using a dummy gage to ensure that the cells tested do not exceed the threshold drag force. Any cells that do not pass this test are reworked and then re-tested until the cell passes. After a rack passes receipt inspection and dry test criteria, the rack can be installed.
Aller the installation of a new rack, the gaps are checked at various locations along each side of the rack at the rack top. Long handled measuring tools and an underwater camera are typically used for this evolution. If the gaps are within the tolerances allowed by the Pool Layout drawing, the rack is acceptable. If the gaps are not acceptable, the rack is re-picked and re-positioned. It should be noted that the gaps represented in the Pool Layout drawing are those that have been analyzed and shown to be acceptable in the seismic, themal, and criticality evaluations.
Question 17 Indicate whether or not you are planning to place an overhead platform on the racks permanently or for temporary storage during the installation of the racks.
Response to Question 17 At this time there are no plans to place a platform on the racks for permanent use during plant operation or during the installation of the racks.
Question 18 Describe the plan and procedure for the post OBE inspection of fuel rack gap configurations.
Response to Question 18 Subsequent to an OBE event at the Vogtle site, VEGP plans to make a careful visual inspection and measurement of the fuel racks to establish (i) whether rack-to-rack gaps j
have changed from their design bases, and (ii) whether the putative locations ofimpact at
{
the top of the racks, which are suggested by the dynamic analysis abstracted in Reference j
1, have sustained any plastic deformation.
I If no rack cellular damage or no rack-to-rack and peripheral rack-to-wall gap changes are found to have occurred, then no further action will be necessary.
l E3-32 l
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If no plastic deformation of the rack cell walls is indicated, but rack-to-rack gaps have changed, SNC will reanalyze the rack arrays with the new gaps. If the analysis indicates.
l that impact loads or displacements exceed the maximum values, then the racks will be I
repositioned using a NUREG-0612 compliant lift rig to achieve the pre-OBE gaps.
l If plastic deformation of the cell walls is indicated, VEGP will resize aberrant storage l
cells to bring the storage cells within design compliance prior to their re-use.
The above summary provides a path of action to deal with all credible scenarios wrought by an OBE event.
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l E3-33
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Ouestion 2 II.1 INPUT SEISMIC MOTION (includesdiskette)
The results of the three-dimensional analysis of HDRs are quite sensitive to the frequency content and duration of the time histories being used for the analysis.
Provide information on how the power spectral densities (PSDs of the proposed synthetic time histories possess reason) ably distributed energy content over a frequency range of interest and generally match the target PSD criteria, in addition to the enveloping criteria of the standard Review Plan 3.7.1.
Guidelines for developing PSD can be found in NUREG/CR-3509, " Power Spectral Density Functions compatible with NRC Regulatory Guide 1.60."
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RESPONSE
NUREG/CR-3509 (Reference 1) describes the generation of a power spectral density (PSD) function of a stationary process in such a way that the sample functions satisfy the NRC Regulatory Guide 1.60 response spectrum.
The PSD function criteria for the ground acceleration time histories (Target Ground PSD Function) are provided in equation 2 of this report with So = 1100 in.8 8
/sec (for 1 g acceleration) recommended in the Conclusions section of this report.
The guidelines provided in this report are intended only for
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ground PSD.
However, based on the target ground PSD function, a conservative assessment of the adequacy of the floor time history PSD can be made.
Figure II.1-1 shows a comparison of the target ground PSD function for SSE (i.e.,
corresponding to 0.2 g) with the PSDs for the Bechtel ground horizontal acceleration time history and the floor horizontal (H1) acceleration time history.
It can be seen that the PSD corresponding to Bechtel ground time history is, in general, an order of magnitude higher than the target ground PSD function in the frequency range of interest, where the floor response spectra exhibit peak amplifications.
This comparison is consistent with Figures 3.1 through 3.14 of NUREG/CR-3480 (Reference 2), where comparisons of PSDs of 14 time histories, obtained from 11 firms in the nuclear industry, with the target ground PSD function are provided.
Figure II.1-1 also shows that the floor time history PSD is significantly above the ground time history PSD.
Both these PSDs were plotted using a three point moving average-.
Because of the oscillatory nature of the PSDs, it is difficult to
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ascertain the relative amplifications.
Therefore, in order to better assess the amplification in the floor time history PSD, the PSDs for the ground and floor time histories were also plotted using 21 and 11 point moving averages, respectively.
i II.1-1
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pg, The higher averaging number for the free field PSD was necessary due to the high frequency content of the free field.
This comparison (Figure II.1-2) shows that the floor time history PSD in the 2 to 5.5 cps range (the amplified region of the floor response spectra, Figure II.1-3) is about 10 times the ground time history PSD.
From 1 to 2 cps and 5.5 to 10 cps, the ratio tapers to about 5.0.
Similar comparisons are provided for the floor SSE-H2, SSE-V, OBE-H1, OBE-H2, and OBE-V time histories in Figures II.1-4 through II.1-8.
Figures II,1-9 through II.1-11 provide SSE-V, OBE-H, and OBE-V floor response spectra.
Considering the order of magnitude conservatism of the ground time history PSD over the target ground PSD, and the amplification of the floor time history PSD over the ground time history PSD without any missing frequency windows in the frequency range of interest, the floor time history possesses adequate distributed energy content.
References 1.
Shinozuka, et al:
Power Spectral Density Functions Compatible with NRC RG 1.60 Response Spectral.
NUREG/CR-3509 March 1984.
2.
David W. Coates, Jr. and David A.
Lappa:
Value/ Impact Assessment for Seismic Design Criteria.
USI A-40, NUREG/CR-3489 August 1984.
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Attachment I d
1
)
The Effect of Liquids on the Dynamic
- n. unm Ceanvas.ag Eng e,
"c=n iiotions of immersed Solids MaasASMC 18 is known that the presence of liquids cen signifcently efus the dynamie motions of.
s imeaerud solids. This paper propeus a merkelJer enelunting juid forcesfor use in t
the dynamie analysis of uning systems in which solid bodies are completely immersi f
Notice
- TlAs instersgl rnay be
- 5*"!"" I'i*"'*** A "'* A '" * *i"' **'**' i' ' * * * **'*# #*"'**
whether e fuid system neay be coesideredfrictionless. The incompressibility require.
Protected by copyright law.
met is aim discusud. Esperimenul deu m eiud u suppers the propend e.enhed.
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(Title 17 U.S. Code)
Fer==8e8/er *> dred ne=de =ssur are 8e6= laded.
i latfeiscties with Ik uid==rimg. and (d) dime==an of analek of multivie-l i
gutnEvts nuliebe move in mutact wille ligeside. Llee ' ;imdum.pw m mth hyde.dpamve e#re is.
)
- hq.nns mmit he di,.pimd w ser..mmmdow th e==ti.
Fiuid Tws4edy Motious With Fluid Compling p sar. m sm r. sed - a r It. Fluid form occur ini th e antids due to the integrated effect of h e preoum. In this Cen.eder h esas of two lung enneentric cylinder, separated paper the came of amvina nulid e cumpletely imenerved in friction. by a Equid annuhm, see Fig.1. The inner cylinder of radius e is nema, ineuenpr ailene liquids in ~_"N.
In this cane, the Ruid surrum ded by an outer enneentree cylindrieel omtainer of inner forse in usunny proportional to the relative arreierations ed h radim 6. De length of & annulus in L where L la much greater amving solida, and therefore give rime to an effective or hydru-then 6. The outer cylinder is nasumed w have a velocity J. and dynamic mann. Where h liquids ausst Aow dynamicauy in h inner cylinder t..
The relative dimpleeement s - z is sei j
tmail paengen, the hydrndynamiie maman naar he many tinies sumed amou ensepared to 6 - s.
A velocity putentis.14 may larger than h solid mannen, even hugh b olisbi many be of le denned (nimuar in lamb (2, p. M] wlm conadora single eyunder large specific gravity. For such nyntems, dynamic anaipes of
"*""h the onlid motissin mint ennaider the presence of h Equide le i
order to provide neanineful resuhs. It b espected ht the re.
Y,=
46 1 64
- ult et this paper would be tvund unefulis b dynamie analysis gi F# = 7 g tI) of nuclear reactor and stenen menorense internens subjected to shore sekmie shoek as well an le Line dynamiie analpes of nonne Auidic devseen, including Suidie & abarbers, r, - radial Guid velocity The eeneept of b h.C7 " onese han twee doncribed I' " ***8'"'I*II"I' "I"*NY by 14tol no it),' I. ands lff; Birther (3), Patton (4), and othern.
Thee reporta have penerally a 1.2 the motion of a single The Suid k onwmiered frietsenlese and to be at reait when b t
cylinders are at rent. Under such nmditinna the Guid is irrota-i isidyin a fluid. In thise paper, exi= ting informatism, particularly fawn 1.amb l:!l, la apr.lini to W dpamic analysia of pretenvi ti. mal and $ will be eingle.vaheed. The baundary condiku are:
I
- ith more than a single. lid nempictcly inui.cr>ed in a liquid.
_M The plau 'of pre entasi,m will he: (a) analvnie ni two.imuly mi>
Or roi #
st r-e (2) theeut with lirluid empimg, (. )Ihenry of multiple.lAuly nmtisnt with a
I"Itsid enupling, (c) experinicutal data on tws> body nmtions N = to eos I at/r=6 (3) 6r The amtinuity equasism in
",Numi'e's in henchets deaignate lieferenssw at eiut at paper.
r 4.I.
-0 H)
- ntrihug3gg key tint l#rmegn blegiHrtrine thWi Mm and pfenented hr hr r Dh' i
d tlat Vehr:stsams (*uuletture. Toennte, t';uuul.'a. Septemtier 1980,
[I""'ipt errewe.t mL ASalE lleadquarters. Jm e
- 8. *f Tese: Awrmicaw Wrsr1rv nr altrunnsc43. Ew.a w ar en..
A form ofnulntiam aan annumed l
- 14. 1971.
l' aper N'* Il*Vabr.100 4 a f(r) cum #
(5) iL hufnal el Entineefint hf leflustry g
f 5
p
-6 i*
To =
- orIstr.ith t*,s + s',n uay 2
...o
,C Frmn egnatinna t7). (B) (12), and (13)
Fn - - Af afi + ( Ali +.if s,)f e (14)
-l s
rn = ( Afi + Af,)f. - (Afi + AI. +,tt,)t.
(l',)
i; l
I shere Fn sud Fn are h nuid reaction intre on h i..ner and f
outer cylitulerm, respectively, and k
Ms = re'l.a = menu of fluid di=placed by the inner I N,'%.
/
u, =.np =,, m a e,,id u.,ad a h.
, c,.
(l eylinder g
/
,e
)
/
lindrical cavity in the almenes of h (16) k inner cylinder I
t i
Ih My = A1. p 4.:
y_,
l d
For the enne of annevntric aphore separated by a frictioniens, in,....; r Male Guid faee Fig.1), the Anid forem ht renuit facen
.l.
esmaau
f a nimilar analysis are abio given by equanima (14) and (1l,) wlwre II 'l Fra and Ffi are the Ruid rescuan forem on the inner se esassau spherm, rompectively, and Fas. I t= b.dy m n
=ase, awed e paso
\\
wk Me 4-re*p = seem of Suid dinplacedbyinneropha "2
s Fran equations (4) and (5)
M.,.
is) v'I' + vf' ~ f = 0 4
We 3*Pr =
s==== af said that co=ld All the outer
- when. e prime indient differentiation with ve. post to r.
IID 8P arienleavity in the sheenseef the h
~k The soleties of this equaden is.
"y st.C'_ ri i=== eP m h
',8'
-a The annissiehis
- v. p + sea m
T v - e.
T*
T, = (- - A) ee. e Systhesis el Fluid Fefcaslof Two Body Probles
\\
(s) re - y/ s + AJ *ia 8 Equatiuns (14) and (15)immy be developed in a mnre neiwral
. o.,
- I,'. 4 way. Consider the case where Guid motion in determined by th M
motion of insenereed solids. Sienilar to Lamb l2, p. IMl, the Suid (9) kinetas energy is taken as a quadratic functios pee
.>. )
8 p _.. (As - 4*)
(le) 27, = AJ8 + A.6.8 + ".. + 24.,t.s. +. "
.i s,e - y (10) s
. i..I A=
er in matris form
[
P - e'
, 19)
J
(
ST =. drJA
,jj Since the velocity of eneh Ruid particle is unsquely detee-f j
tnined by b generalized vanables of metana si and z., them where k is a =ta-vector and A is a square matrix. Siam any quadratic form een be esprosed in terms of a symmetrie m A frie*
l==
in, IAgranse's equations of motion wiu apply.
. ' '. I Suid is tre a-a'---' unstansthEtendal emerry_ n"d mar he gallsten jpentini Lt
- - ~ e& '~flio [5] the snans matas A may be sonsadered symm
/ n A,
Forthetwo.bodyproblema 3'f
Guid reactaea fesus F is seek a syntesa is given by Lagrange's 2T = 4J/ + 2A J.t. + Ai.t.*,/
i f
(20) equatius f
Psv - ~7 g,g + bT i hT j
(11) Fras equations (12) sad (20) 3,,
F s " " A n#8 - d'.t.
(21 i
wlere s, are the generalized coordinates of motion and T,is the
_ j g,.
gg In this poper r, will generally be the tra >
Auld kinetic eneigy.
lationni nuition of a solid body (body i), and Tf, will be the Guid and Ff are the Anid renetinn forcea on w
~
where menin, Fn
/ reactiim force on that solid Inde-' It be rennonable to neglect the contribudon of the last term
.'c Tlie ruefirients A mill nuv lie determinei budies I and 2.
equation (II)if W solid motions are assumed to be small with wme for this example that in Such r.n maaumption is w h eimdition of the pioblem above for the ryhnders an, rompect to Buid chaemet thicknenen.
Now equabs (21) and (22) are genernlly tine int i h
made in this paper. Lamb (2l considers a few enwe of single-sp h body motion including the last term. Negieenng the last terra values of f. and te. If f = s, en t of equation (11), W fluid reaction force is at every ru. int in an inenmprensible fluial and a pr., ure grada asiste througlwiut the fluiJ due to the finid inertia, l
g (approximately)
(12),
ble
(.,
F, = - di g = ps, l
f e
t The finid kinetic energy is Transactions of the AS
}:
1
..t Tid prn ure di.trdmtion gnsi re,e to a buoyancy f.res of all t'a are eitusi.
F..r thi. c.,nditani the Cnid f.rie arr u-nnlly an Archies.cdc type, ove ht enady descesninside. emd r ta & tw..-l.dy e sp.- alici.dy de,
- "'",hed involving t he enmen.enta iA A.
"I' I"""
" ' " " * " " ' ^ ' ' " " " " * " " ' ' " " '
F i = -(A n + A n)s - Mis.
(24)
- d f
e estaldi There are afn +
l f, = -(A n + A nis - - Mos.
(25) IU2 cuenpments of A shich mies be determe el. T bc re-f s
f un. which maisiing equations may he establi%hed ley netting all s, = 0 ca.
cept one, sj, and lettinc j = 1 to n.
The vahwa of & Guid Aus+ Ass = -Me (26) fornen are munt easily determined if ordy nne laady at any eme time in allowed to move. It in suagented ht these Auid foran n + A n - Me (27) 8"d would he determined from the ein tinuity of G..w and by une of equation (12). Altlumah it ia dinicult in predict all the insnible Me = b mass of Auld di placed by A inner body conficumtiusui et may le net in practies, it in augen ted that in u.tvisig h amtinuity cepatirm, enne enetim.d of mariam and M, - the maams of Auid ht unuld All h budy 2 in the alw parallel llow impedancea miaht he ense.idered. ansluguien in an ernes of the inner Indy electric network analyain. Fulhowing thin primeriptime, slie ann-P""*"'" 'd
- 8 "Id ""*" ** * "* d I" "'" # ** d"~
Equatm' em (26) and (27) pn wide two relations. To evaluate I""'"*"
the three unkiiown Au, A,., An, a third relation is needed.
Annume the rimtsining landy 2 to be statie, s. - O.
From P""'"" "'""
N""
minat.um (20 For the multiple hudy problem the analynt nuiy And it nw.re eenvenient to a thedae h dynamic problein by numidering Ts = - A ns =' - M,A.
(deAnes M,)
(28) the espanne ed eingle channeln. The preneure distrdmtam of thee chaemels een be writtai in tersen of entrame and cut Auld As indicated, equation (28) deAnas time tonn M,.
M, anny be veineitisi and channel wou ow.tione determined by the samtion evolunte4 by annundng h body a to have a velvcity s. and by of isomermed anlish. !!y emelderms sentimmty and numientum the em tm, uity of Sow, the Suid valueity distribution may alan be or sentinuity and Lagrange's erpiation, e merieg uf espeak.a re-synluated. The Suid force may be evalisated taming the 7 Je suit. The premure distritmtiueen are then armaidered an elenwinto tien of mamentusa or by uaing equatase (12k which residte in of dynasnee fores generation in the equationa ad messun of b 2r enlids-An eigenvahne Problem residta, wliich can be es6ved to Ms a d' (for de = 0)
(29) develop the solution for frequeney and deAertiunal reigsmises, 8
given the assumery '
ig sendstiuna. In smany eems, it where T, b the Guid kinetie energy. Eines the sneenantum rain, nasy be nerosary for a Said speciales to work with the dynenues tien will give the Said promine which enust be integrated to eb.
epocialet to develop the enhetion for the enenples Sand sehd tain a Suid faree en an immermed budy, the see of equasism (29) problems.
(
will sammally le maapier. Frani equesteene (2th (22k (28k (27),
In name teams, Guid e.m;- 7 acts in omienstium with k.
and (25), equations (14) and (lin) follow. This, these reinteens Said inertienes or hydrudynasnic men to raune frettuency anudse
) which more derived from besie 4mid -- b-are abe obtaan. largely due to the Auid. An esample b a licimlads roumater i
able by the mehd of ernthmen desenbod almve.
forsned by a tubehmet vibrating relative to an adjacent pienvin.
g The deta in Tabis a are typient af assis avadable infonnataan 0=alistiun of the tubenheet meet be meanspanied by di plare-r giving hydrudynansie naam relatiuan where a single body is in ment of the Suid. Thb di ;'. _
t een be arnanniudeted f
nestion and in nurrumuled eitlwr by an unbounded Aisid initially by the. g "ty of b adjesent Suid and by the Auw at rest or by a static sentainer. By une of the above prouwdere, tbnmsk the tube whink came an inertiente e#ect. The author these tabulated data are tronatorniable into hydrodynance snain LJ J equatiese af motion far such syntenw by Arwt emundng l
reistione a here the single body is either carrounded by a mnems ht compromubility won su l.,w ht only. tube Ainr was ism.
sentainer whune dimennionn are large earnpared to the singis portant. The snethuds of analynia of thb paper were then ie ed e
I body for the sense where a ningle body is shown in Talde 1 oc to devoluy dynanme eqisations. 'Ilism. slw tube Aow impedasse where the outer nurface for Camme 8,9, IS,11, and 14 may be was asumed to ho so high ht --, _ Nty e#eets pro-f seasidered in amnion.
dessanted. la such a enae it was easy to write the equatsie i
The reader may note in equataan (14) that when is - se, the for a Suid opnas. It was then an easy step to write a ense.
hydrodynande inams Me in b dimpleend unami nasang fross busy, tinuity requiresnant osanalering both e5ecta. Another example r
eney. Tlie hydrodysiencia sness M, in equansen (14) is ansaan. of euela a IIstmbols reminator which interneta with the neerhanical tied with relative matium and stay be emnandered an inertal systoni eseme in the analypin of vishna, a
j equeese.Aln eteet. In seassak the b, M, -- S amass een be mmmdered to anwet of thus besynney and inertial equeno-Elm esoponenia.
Tlle Eteet el Fluid Damping
~
The preeeding analynis monumes a friction 1 ems Suid. L Melliple Body Metless WRt Fluid Cosping ensinanag d gner nu.6 have some guidance ta judge when a i
Where many badies see imenersed in a frictionima,inamsprem. Guid may he runaidered frictionios. Sosne guidanee in permented
(
ible liquid and are coupled by thin liquid, h nectived of synthenis here.
I diseribed above for the twie.liudy pn.bleen should facilitate h D I'i'1I""*I P"ure drop is annunned to be based un the determinat ion ot h hydnnlynandr fones. Thin metim diseum-Darey friction factore, obtained from eteady-now data, marired for ihe multiple imdy problesa. From equatimes (19)and (12)
AP = fr, ye D, 2 p (31)
-F = As (10) f whom where F and i are (= X I) aduinn vectors and A is a square f
rymmetric matriz (n X n).
F, is the vector component of nuid AP = frictional premure drop Iarre em ench maid twidy
- here i,in & instantanonum acceleration f
I>a.cy friction fartor
- I that =. id laidy. Tu determme the comimments of A, we ob-I,
. lenctis of channel heryg ghgg gjugg pgggggjgg (;lQ) ggp g hg gggggdjy gngg, jg ygg,g y
- ObMmici velucity,82Hmed Unkfprm kn Cbannel e;pply for any georralissd input vector s.
We an.ume firmt tiu.t p Guid n man denaity Josinal of Entineerint for Industry
- - - -...., tro 3
I 6'
)
~
t eu, s seyd,w,.e 4.
... p.sc 3 s.n
..,... t a, sa 61. se i. w
.d es se.a c.a.
u.
.e m.,..n se.
6 4
... i.,;.
t t.mp.eed e. ena g6,4. t.dy;.s.in despe.steet e.
,3 ss es..sfl I
t t
{
al
', a Ditestsas et trer.apemmes ans.
f.
I lhe.
St..ruen <
tr.r pemake 8hs.
- Ast, gem emL6e
.'m.. pg.e.
fre.8, t a.c.
a
- s. s c,ne.,..
- 4.
s s
.,u j
R.
I 3
q) 1 n
.. ~. -
t, I
t I
- s. amat msaws
,- 4.
- s. m t
, r..
y
.h Le ammmans e an6.
88 ten F&ar
,,g,,
8 1
.nsmen amt..
- /
IS 1.34 Obe. emmmans.aeeamroes anna &ar ab 4-h.
3 1
1.54
- a L
1/t 1.18 8
P,.iast. 6susa.
T.ree. ores 6.%,g h L. en n/5 1.W a
p.
A/as 2.7,
$ **i-i a=<
6.
1; f.
stame e e.a e 96, asurenset. f essen. eassas 6
est.a. e n. .nsmer tem man Y
- $ {F ge hamme Asun&g.
Esep.
n/b 5.
g, per es.
t.rs..amun.6ama& a,sa.kam af n.r. a = v..
g' 5' * #
-T e
A 6 c.
u.. - >
=
u.
r er u
as m
J
- k. -
A/5 g,ggggg game
-n <
.. am*. senne,. :h>.
1
.l n ammm.as assei., a >>.
5, P.,..
. c-
+ c-8
.P.*
m* \\
L. fob 5 m
- ' o ab.)
. set e
6 ad "
6.as 9Ah i
I,)
3.as
.se
.e-R$.
&fF4 888W p
1 1
2-
',y
. 6.,
~
2 f
- 6. he.asaguear Fl.ts.
Suunt 4. Pha hAb E.
- f.
~
~
.N',': *
- e
'Q 8
M g *> ;>.
9 1.8E8
./
n.am r.'#k':ad 4!.
s.. vi r #.,
g
.u,,
i j
nh mment I.urnama I
)' I e
1,e8
.3ES
.9ES 95 m.
B.5 4
.1 g,
s.an
.am
.a i.as
.ese
.ssi
.)
T as Ma=*=
D, - hydraulis diameter - 2e for a channel with paraDel Amun.e the finid velocity to Ise cyclie, walls and separatinei thickness e, which wiu be used 1, = Y. s.m wt (33) ha this exposition
' Using equationis (33), equation (32) may be integrated over a The frictional energy II half-oile to give Er = JLoAY.e (34) t' Er = gl APAYes s 3ue (32)
If the iniid damping force were linicar and equal in 6Y, theti pj the energy Esover a half-cycle is Pade
=
4e 6F'Je 6F.
.in' utJwa Ks =
ww
.< o
.o A = lluial area v 6 P.*
p
~
2w
- = tinic v......et....# e i.. a e lA F
,n The effec tive iment damping cen.fbrient I, snay be determined enn d to acew rsaiily with nedh t um.
Foruilwrt=nntauth's i
f 1.y nettii. Eg = E,, mith the re ult graplwd data, the vermium betmern n cnamed nintm ni ficquence am) predwted nat m al fre piency
- aa is parally lev. th.us 2 percent.
]
I 2 II'*8bd I
g 3r y
Fier b = 1.2 euul a in tuency ul 29 eps, the masimum vahie of tle lleym,hi, number durma the vibratory cy le in e timated
{
e
" * "" **I A parnmeter 2( may be defmed whers ta occur if the Itr> nolds immber in gienter than 3000. Thus, the water surrounding the beam ran be ennsidercrl turbulent, e
II " Ag (37)
Itclathm (30) give,. a value of 2( of 0.0:1, using a liietion factor I
of 0.027i. Thia frictism factar is taken from Moudy [7]. Sinos
}
and Aff = fluid mm.m.
2(in the reno of the dampin,t impedance 2( = 0.03 in much i,maller than 1, the ameept of thia damping
?
to the inertial impedanm. ( is similar to a fractim of entical parameter would imply that h fluid couki he consulered emen-d2mping for a one degne.of.frecriam eyntem with linear damp-tially f rienonnena. The fact that the data on natural frequency l
ing 6, mana Af and natural frequency w.
From equanons (36) agreed no well mith theory wimid tend u. validate the amumption i
r sad (37) and aith Af = pLA.
of a frirthmican Auid. W amphficatism in Keane's to t was f
about 15 at ro maure, which can al o be connadered an evidenes b
2( =
(38) that h overall inertial impedance in considerably greater than 3r we h overall dampmg impedance.
Fcarhr,if F. = use, from squation (38)
- '* 'd es ead Kin. Frits and Ki Pl reported the resulta of a tot on a alid aluminum cylinder Aeribly supported within s 2 fze rigid rylindriJai omtainer. The equipment was vibrated on a I~Ee Gurbulent now)
M shake. table. The length-to. diameter of the eyhnder was.about, 8
1.0.
The cylinder wan surnnanded by a thin annular Auid which f
Equata.urs (39) definen a dimennionless number, a da.anpmf won free km Auw axially an well an cirrumferendally. The natural parameter, that should proviale a renaonable measure of the ratin freguncy waa taken an the frequency at which the vibrational af Auid friedon to Anid inerna. We reenin ht in equation (39) amplitude of the cylinder reaclied its maximum value with a enn-
{
f = h Darcy friction factor for turbulent now staat table amplitude. The saial and circumferendal hydro-
- s. = h dintanes that the Auid moves is an a ditatory cycle dynamic maa<en wm osmhined an ohuwn in itesa 10 of Tame 1.
g, (asapiitude of sinusoidal mouse)
The natural frequency of h cylinder in air wee 35.5 epa. With s - the Suid ehmanelopeans water _._ " ; h eyhnder, the inquency wan reduesd to 17.0 eps, which gave very estinfactory agreement with the pre A similar analysis for lasmanar Sow throusk a parallel plato diction of 16.9 eps.
chamael gives the result In reference 18] h Reyne&ds number was estiinated to be 4400 whiek wea sunsidered turbulent. The value of 2(la cateu-t 2( = b (lamusme Sow)
(40) lated from equation (381) to be 0.03, using the data from reference t
/
[81: f = 0.04, F. - 3.2 fpi, u = 2r (17 epa), e = 0.000 m.
4 g,
where Sines 2( in much less than I, h amumption of a frictionless 4
g Sund abound be valid. This is validated in that the use of the 3
F = Ouid kinematic viscosity bydygdyyamic masa comgpt in reference l8] provided a Tefy w = angular frequency of oscillatory motion e - Suid channelepacing
\\s If the concept of this damping parameter 2( is remannable, i
i then the assumpdon of a frictionless Suid must require that 2(
7" i
must be much smaller than 1.
The quantity 2( eill later be r
ealculated for some test cassa.
. :9 1
[
i n
var Fluid Compressibility
/Iuid arrutus:
q)
The preceding analysie assummed na in_
9 Suid.
of dgtf C
\\
i f,
Where h pannihility of a Suid sprms is preemet, it is usually a q
I I
straightform ard calculation to deteressee if h volusse storage of a Suid spring wdl atest the emotinuity han====
The application gk i
of this paper is further resaneted to esses of small Mach number gjord q) l l
(less than about 10 percent) and amnes where the Sow channel 5
t length is amall compared to tise wave length for propagating l
l vibratory disturbances (less than about 10 percent), in order to
[
svoid the prmaibility of standing-wave effecta.
g g,g jj l
S- --
l Comparison to Test Data 6
~
d
._o Ke==e's t>.ee.
Keane 161 vibrated a circular cantilever tube l
currounded by an annular envity. The empty space was fdled I
@,i eith a liquid, a hich made the ten nmaintent with item 8 of Table q
l I (due originally to Stokee ll]). Kcano esamined Aczural vibrati.ma and analysed the emntdever beam with h added g-h, l hydrodynamic mana. For one set of testa Keane's beam length-i tadiameter raGo was 17 and 6/u (in ild m A. Table 1) ranced
! riti hW p
from 1.2 to 61. According to Fig. 4.3.2 of Ecano's report, the hydrodynamic mans reduced thu natural facquency from.'.G eim (air in cavity) to as low as 29 elm (with water in ravity), miuch Fie.a ve*..
- e..eeg.4eamd 5
8 o
f.a43 2 Ceteves* a. 8 21, 6se.ea. mi s, nouse e..,..ae a..a scrurste **nmste of the naturst in. inn.cy ii sl.e pre.rure of la es. 3 o
the laptid, A roncentric r)hmier aa-An aluminum %7--
=
'a,
,, 3 [.;.,,
~~'
h Vib,.ee 8. C.a..asvi. Cr u
ecmbly *lauen in l'ig. 2
- as available f..r te=t.
~ ~
cyhreder (part fi) na flesibly supi..ried by cailumn etruta (parts I'luid lenkage L
I). The cyhnder in marrounded by an munulu..
~~
from the annulus in limited by ch -clearance metal meals, parta Tlie motius of h rylinder was measured by use of a canti-
~"'"~""
4.
The 5
lever di placement smee. fitted with strain sages, part ti.
a
=
=
=
=
-d
=
=
=
fluid annuhn. width e was varied by machining the cylmder diameter. Table 2 simwn some numerical data frum theme tewita.
g,
)g Fig. 3 ala>=s some typical cacink staph recurds of the freu vibra,. * % "-* *"""4
,, *7' *" " *~*'--
'f
[v tivos takas during tho.e teate. The vibrations with mir and the
,,,_,,,,,4*,g*,,,,_,
glycerol welution were r.reated by velocity.*horking the cylinder i
f,,)
with a large mallet. The vibrations with water and siil were g,,,, y f3 ernted by caming an initial large dcAection.
The hydrodynamic weight was calculated from item it. Table g g g g g,,,, g,,gg g ggy g g
.' s Theme equations for two.hudy nmtionn presented in Ij The hydrodynamic weight was dwa calculated from the test paper.
IJue to h natore of W equipment it was felt this paper conneet buoyancy and inertial aquecee. film dects.
1.
- y frequenn,oi.
that t ere was a small amount of leakage past b end seals Therefore, references 161 and lB1 are conni tent wi
,.; p h
In referen:ee (6) and (R) h buoy.
4{,
which w. wid enune the actual hydrudynanue weight to be smaller presented in b paper.
than tie theoretical value. Even Osuush the value of 2( was as nacy term implied negligibk efects on natural f.e y
"1"liis
. (..; g high an 0.7, de calrulated hydrodynamic weight was felt to be imply a 4ssmt de a h prdieud WiMim.
agrament with es t t vatum.
.se = minpiismiusi w epecisedly riuwd ii referew pl.
- c is._-
,,3,,,,,,
It a. noi.d 4.s s.ppem. wm ai u used a om.-re se
%,,,,,,,,3,a,;,,,,,,,,,,,,,,,,,,,,,,,Mah
- [; z.- )r mdels k ein papw. Adnudin m hydrudynaanic mam of a thin annulun around a rotating cylinder.
,y; The rmdto wero pubE,. lied ia till' is desirehle. )(owever, eines the inforenation of this paper in
. e - ri.e -,,,,,,
f.c,'l ;
m
==powe= to ist data app 67 to the dat d the inertal equem g,,,,,,g g,, g,,,,,,,,,,a,,g;g;,,,,
uj:.
9 ska n.d de me et me dumping.-.m st. no /
er the i msia eg sie er, in.the w d., es virumi Comosets of listi Degiss et Fresdes Dysamis Analysis esses, w an,t pr.d.ewa by a.k til, aw 6. ei
- bem, A esiians inte. d the mest =deir med male ' _ _e
- i-
=dmir samspud. Thessa., eonarmaim. et this esat is met freedom dynasmee analyses of linear syntesem =Inick me, P p:
llowever, both rderesses (61 and (81 involve semular by esmo arbstrary home inutauen is the tramform I
now.
Amid spans where the outer and mner buundaries of the auid trary sentsuratium of dynamie cumpiments into i!
ennulus both move. Boe refereness (6l and ($1 report forsee neray d simple serillators which are eseited by h base
.i*
of Suid motion equations (et equation (4.2.13) of lal and sque-
, y;
(
5.
7.. g. p..\\ A \\ \\ \\. \\ j
,, it
.gff.
r..
.,..;,.f,q L.f
./.,1 - -l j
?\\ ~}.W.U \\W T.\\ A 1 l-
- ; j.
_ g. g i. 3 brt 4 i.
g
...6 l
i t
i-
.i.;
f.~ x
-i-
. j-i l
p M h h
-tj ki fl h
. I : l:gl.-1 W I. ;
~
. 'i
_ p. \\. s. q \\
l f. ;-]:.VJ I %. / I I i l
.\\ \\
,6tr-m-
..a
+
- -y-
- ;-; y t :rl Ll'"' [_l__I I.
.i4
+m
-l tl.
a
- r. ; -
y
.ossw w,1
>s-I.
i.Ie I~
'N I2 i!
f, 3--l, -.f ~3 F,I gl i
l j
pi
- e. y ; q !
51. if IId M-b O,,.j
~
.l.
- { 9[ 1 1. j
{
it-I
-l! i r i. i. g1id.
i W W4W H t i-
.o M h WW \\
._y
- t. \\ U \\ 4. u c i p ;i.. m - L.2 r;T T u\\ d.d-\\.\\.
.\\
ga \\y&M y s;g 1 U
-ar-=-
e,,,,,, _ e.
g
)
e,.,a....
.., w.
e o..i..
a.. c - e.as 4 n
I Trsncirlinnt nf the M E--___-_ - _ _ _ _
n I
Tl.c.iu.tur, of ibe coing.lcs array a thru relate d t.. the nw.tiam of 3
!.=rnh. ll., fly + a.Iv=ea.dre. 4,th eJ.. l >os sr.19 4't. ch. t..
l the mnple. em dinters..\\lany foruvi of thi. tran-formatu.n
- n itieu. oft. C.. Nw.fredene. ave. A % fy *= fnie. En,e a.id bnvr hewn rep rini in the literature in equati.na t hich are in
- '"' J d "d' P' ""** U "'"'' F P'"*
I M t h 4 rament tw., applwable mitle.ut dynamic nnipling, that is, for the 4 Patten. K. T "Tal.len s.f Ilobudenamer Ma F,. tore for
. d
^
enee alere the sunne snatris i= disgunalin the dynasrae ertuatiunn.
5 Creadall.11 and McCalley. It. D A.i.merical Mal Je af The enrrect trnesformations to be inned for the came of d)mamic As aW. Eart s'id fdeetwa //esulM. Vol. 2 (ed. C. M.11 arras coupling, ml. eve the mann matrix is nondiagonal, in given by and C. E. Credeh McGeme.Ilill. New York. N. Y.
h!cCalley in referenn (9). The methods of hydrodynamic G Kune. J. A.,"On The Elati.: Whrstion of A Circular Canti.
analyaia of thi. paper generally remult in dynarrue coupling and lever Tube in a Newtoman fluid." PhD themie. Carnegie Institute should therefore u,. hicCa!!cy's relations (or equivalent) when
',f Technolosy. Sept.1963.
treating fluid a ficcia in a multi-degreimf. freedom analysis.
7 Moody. L. F Fnction Factors for Pipe Flow." Taana, j
ASME. Vol. So,1944 pp. 671-664.
8 Frits. It. J.. and IUse. E., "The Vibration Responw of a Canti.
$E5ffl3ff levered Cylmder Surrounded by An Annular Fluid" KAILM.
G539. Feb.1966. (A radable from Clearinghouse for Federal ScientaAe Some availalde relatinne are given in Table i for hydrodynasru.e and Technical Inferination. U. E. Department of Commerce. Sprins-mannen for nuitinna of a ningle sunlid bndy fully immerned in a Geld.va 22151J frictiamle a incumpre ble Auid. Thin paper proprv.wi a method e McCalley. R. B " shock Analreis by Matsu Method. Notes of u*ing thew it,ulta for tuu-body motions. Where a single for Short Courus e,n Normal Modes.".Wek and Funetian. Depart-body in alu.au in Table 1. the semnd body is amsidered larse was of Engineenns Mechanica, rennaylvania state Univernsy.
compared to the einsle budy. For camme a, 9,10,11, and 14 the outer surfan,i may be evnmidored in arbitrary motion. Some to Nvete nwnanimaenties: itesna 12.13 are due to J.II. Germer.
General Electne Co.
guidelinen are pn.pu.ed to establish the conditions of frictaunlemn, 7,,,
g, y g,, 3,,,g,, 73,;,3,, g y47,,,
incomprewbie Raw. The came of motions of multiple imrnenied tiene of a long iteter. Part 1-Thouer: and Part 3-Test." Jewrnal ochde is con idered. Compartoons to tant data indicated favor-of Aaes, gn,in,enne. Taame. ASME. Senes D. Vol. 92. No. 4. Dec.
able agreement.
1970.pp.923-937.
13 Kina. E., "Anninie of the Fundarnental Vibration Frequency References d * "*did V""* l*d 8" C'"**'*' at,.ct
." A m ess.
Pressedines e/ Cesvarense en flew-Indused vihresione in Resanse I stekee. G. C "On Snme Ceems el Fluid Motion." 73esafimee Sysseus Cempements. May 1970. Arename Natisant ' " m,.
Osm6ndee railesepairal See Vol. 8. May 1848, pp.10hl37.
Aryenne.15.
Pt.
S t
I r
b O
P 1
l
[
t.
g I
i I.'
a j
f n
1
I..+.--._:
Group 2 Question 6 DYNAMIC COUPLING IN A CLOSELY SPACED TWO-BODY SYSTEM VIBRATING IN A LIQUID MEDIUM:
THE CASE OF FUEL RACKS
P OC. OF THE THIRD CONFERENCE ON "VIERATIOJ IU NUCLEAR PLANT *, - 1982 BRITISH NUCLEAR ENERGY SOCIETY (1983)
DYNAMIC COUPLING IN A CLOSELY SPACED TNO-BODY SYSTEM VIBRATING IN A LIQUID MEDIUM: THE CASE OF FUEL RACKS A.I. SOLER Vice President, Holtec International Cherry Hill, New Jersey, USA K.P. SINGH President, Holtec International i
Cherry Hill, New Jersey, USA
'SUtCIARY An approximate analysis of the effects of confined fluid on the c: ass and damping present in high density spent fuel storage racks is perfor=ed.
shown that inclusion of large displacement effects is required to Ityielt. realistic results for rack forces and pool flocr slab forces.
The theory is developed for square cell geometrias,'an:i n zieple two degree of freedom numerical example io presented to illustrate the effects.
~
NOMENCLATURE
)
Characteristic dimension of gap (Fig. 2) c f'
s Friction coefficient (Eq.16)
Nominal gap between body 1 and body 2 (Fig. 2) h Cap in annulus i at time t. (Fig. 1) h i
K Loss coefficient Er Kinetic energy of the fluid set in motion Length of bodies 1 and 2 (dimension perpendicular to the L
. plans of motion) pi Ilydrostatic pressure in gap i.
Generalized forces corresponding to System Lagrangian in X Ofi,0f2 and Y directions, respectively.
815
g *.
Length co-ordinata in gap i.
3.
1 Uj Displacement of body.1: (inner body) in X-direction (Fig.
2)
Displacement of body 1 (inner body) in Y-direction (Fig.
U2 2) s Displacement of body 2: in I-direction (Fig. 2)
V1
. Displacement of body 2 in Y-direction (Fig. 2)
V2 8
vg Width of gap i at tima t (Eq. 3)
(,$.
Velocity of body inner and outer bodies, respectively.
eg Dimensionless width,of gap i P
s' Ratio of 2.ength to nominal width of gap (Eq. 7)
Hass density of the fluid medium.
p INTRODUCTION 1.
Dynamic coupling between proximate bodies ' vibrating in a fluid medium is a well known phenomenon [1].
In the special case of a two body system executing planar motion where one body is completely enveloped by another, Fritz (2) derived expressions for., the virtual mass and the couplant I
inertial force under ths-enemmetden that the. motions are of small amplitude.
Frit='s war't ~tras ' teen te7 asis or the dynamic modelling of fluid coupling in much of the structural analysis performed in the nuclear industry [33.
Dong (43 sives a concise account of previously published work on dynamic coupling between closely, spaced bodies executing small amplitude motions.
Unfortunately, the assumption of infinitesimal 4
vibrations is rather untenable in many applications.
The case of
" poisoned"# fuel storage racks containing spent nuclear-fuel is one such example..The fuel rack dynamic coupling problem will be described in some l
detail since it provided the primary impetus for this study.
10 isotope, used'.
The term " poison" denotes a product containing the B for capturing neutr.ons.
2.
A " poisoned" or high density storage rack is essentially an assemblage of cellular members of square cross sectional openings. Fuel storage racks are about 16' (4.88 m.) high and vertically submerged in fuel pools containing approximately 40' of water. Spent fuel assemblies, after their removal from the reactor core, are placed in these cellular locations for j
long term radioactive decay.
3 Figure 1 shows a typical channelled BWR ' fuel assembly in a storage cell.
Curbs on~ fuel reprocessing have necentuated utilit'ies' need to use l
poisoned" storace racks, as cpposed to open lattice construction employed' in the past.
Water in the pool acts to modcrate the emitted neutrons and to transport the spent fuel decay heat.
However, in the event of an earthquakc, it also produces dynamic coupling between the fuel assembly and 4
e Atre
,., tha cintain*,r is in tha ordse of 0.006-0.000 a. en cach sMc. [Per exanpSe, y
p3 csils for min funi assemblics (approximately 0.133 e side dirensien) are typicsily made 0.1524 m squsre.
When subj:cted to a civcn grcund gotion, o fusl assembly is frem to vibrate end local impacts with th3 storage cell may occur.
The magnitude of impact, of course, is a strong function of the dynamic coupling between the vibrating fuel assembly and its surrounding vibrating fuel cell.
A multitude of fuel vibrating in unison and impacting storage containers can yield a large overturning force on the storage racks.
Ultimately, this load must be borne' by the pool floor slab, and its supporting structure. Moreover, in free-standing racks (unanchored to the floor), excessive rigid body displacement of the rack, and consequent inter-rack impact are also areas of concern.
These considerations indicate the importance of modelli:g the fuel assembly / cell dynamic coupling in an appropriate manner. This is especially important in operating plants, since strengthening of their pool floor and support structure is all but i=possible in most cases.
It is i=portant to develop a' seismic model that yields conservative results for flocr slab forces, yet is not so conservative that unrealistically high floor leads are obtained.
4.
It is recognised that the velocity of water in the gap between the fuel assembly and tha storage cell will be three dimensicsal. However, the axial component will be quite small since the length of the fuel assembly is an order of magnitude greater than the characteristic gap dimension.
Thus it is sufficiently accurate to model the problem in two dimensions as shown in Figure 2.
The outer body of square cross sectional opening simulates a single storage cell.
The fuel assembly is modelled as an unperforated. square cross-section to simulate a channeled BWR fuel assembly. It it intuitively obvious that the effect of elastic deformation of the cell and channel' walls on the fluid motion will be insignificant.
Therefore, the walls of the. two bodies are assumed to be ris;1d.
For analysis, the inner body is lutelled ts -body 1, and the outer body is labelled as body 2.
The fluid inertia forces on body 1 due to an imposed two dimensional motion on body 2 are determined by writing the system Lagrangian for the fluid kinetic energy assuming inviscid flow. Lagrange's equations of motion are used to determine the generalized force.
The amplitude of oscillations of the inner body (body 1) is allowed to be ccmparable in magnitude to the inter-body gap.
The 'resulting expression shows that the dynamic ceupling consists of a virtual mass, a coupling inertial mass of the type derived by Frits, and an additional non-linear force which may be referred to as " fluid spring".
These three forces completely characterize the fluid forces for large amplitude motion under inviscid conditions.
Fluid damping due to duct flow type losses due to form drag ca'n also be derived from force equilibrium, if the duct turnaround coefficients are known.
Expressions for equivalent da= ping due to drag are also given here for the sake of completeness.
5.
In the following section, the detailed analysis of the subject problem is given.
The results of the analysis are illustrated by treating a typical numerical problem in some detail.
The primary intent of this analysis is to demonstrate that accurato determination of fluid effects requires inclusion of the effect of motions that are large compared to the gap between fuel assembly and cell walls.
To illustrate the potential effects, a simple two degree of freedom model is subjected to a simulate'd comparison between small and large displacement seismic event, and a l
817 L _-___
.salutions is ends to illustrato cerors sustain:d in using the small dispiscs: rant thscry.
gproximateAnalysisofDuctFlow 6.
We consider the 2-D cross section shown in Figure 2.
Neglecting out-of-plane flow, and assuming small gap width hi compared to the characteristic length c, permits the incompressible flow continuity equation to be written, in each portion of the duct, as 5
i_
(1) du,
=
k Di which yields the solution for the local fluid velocity
( e+ - e ) S g+Bg(t)
(2)
~
Ug=
3, 1
In terms of displacements Ut (t), Vi (t), i = 1, 2 shown in Figure 2, we define i = W (t) / h h ; i =
1', 2 (3)
Wi=Ut-V; e
i i
If h is defined as the nominal gap width around the entire periphery, then the current gap widths in any portion of the cross section are given by the relations h,3 = h (17 Ej) ; h,4 = h (1 ~E)
(4) 1 2
2
, Applying Equation (2) tavsactr av,gicq,3 and.;. int.e:Tpating Si as either x or yi Z, = - Ui, Z_ = t V, yields 1
O = j /h;,+ B (t) ; u2 * -
x/h2 + B (t) g y
j 2
2 (5)
' b
- 2x/hg + B (t) 0*1Y/h3 + 3 (t) I 4
4 3
3 direction and impose At corners, we require flow. continuity in the local Si the following conditions to determine B (t);
1 U3 (c/2) hg ug (c/2) = h2 u2( /2) ; h2 u2( 2)= h3 (6)
U (_e/2) h3 "3( /2) = hg Ug(_c/2); h4 ug(c/2) = bg g
where ui (c/2) denotes ut evaluated at S =h in region 1.
7.
The following final solutions are obtained for the approximate flow velocity distribution.
~ y$
y$
2 1
Y/
- y=
/
U (S,t)
=
+
g h
2 (1-cy)
(1-c )
y (7)
UW pW 1
2 x
j U (S,t)
=
y 2 (1-c )
(1-c I 2
2 VA VA 1
\\
U (S,t) =
+
Yj 3
c y
l 2 (1+c )
(1+ci) y 818 L_
pA 90 1
- /
U4 (S, t)
=
I 2 (1+c }
I1+C2 2
The kinetic energy of the fluid can be written in the form C
4
/2 fpb L
(8) y g,
,g gI)
Sdg c
g i"1 ~f 2
~
S /c and L where we have temporarily employed the notation ui e at + gi 1
is the length of the container. integrating yields j
p h.
2 8.2 '
4 CL (9) i
+ 12 f"
2 i=1 *pi y inspection of Equation (7), we obtain b
After eliminating ai,
~
3
- 2+4
- 2 ocL (10) 4 Xf" 8h 1
1 2
2 with 2
2
+
' 1" 2
2 3 (1-cy)
(1-c I
2 (11) 2 2
=
3 (1-c I
I1~'l 2
We fom d [I f-f
~
5
)
O g = site k'1 Wg)
- 3. W g
g and obtain the formal results f
,. 8 +1
_ y 3
oC g-84 3
1 Q
fl oC g
L 4h 1
1, 4 h 1aW 1
3W y
2 2 + _I 4 3' I 4 2
2 1
oC 8h aW l
8W 2
~
1 1
3, h$
84 2 3
2 O
~
f2 C
pC L
4h 2 2+'4T 2
8W 1+3W 2j N
y 2
2
- 2
- 2+3w pC 1
1 2
' llT (3 W2 2
8.
Using Equation (11) to evaluate 04 /JWi permits us to finally write 1
Qri as i
$1
- -2
- $.2 "1 "2 2
"2 1
2 a
a
~
W+
Ogy.= my y
6 1
l (13) i
)
@2 2
+$
W 2
2
^
+
~
O
- "2 "2 +
6 "2'
1 1 "2-2 "I
~
f2 where
'I OC D
'l 2
(1-c 2) 2 -
i ",1,2 (14)
I
~
1"h y
3 i=1,j=2 oC 3
g y
+
(15) 2 2
i=2,j=1 2h 3(1-cg) gy_,j )
l 9,'
Equations (13) and (15) provide the' contribution of the fluid in the -
to the system Lagrangian and includes all conservative effects.
We gap clearly see the virtual mass effects in the leading terms of Equation (13).
The coupled quadratic terms in the generalized velocities appear solely from the assumption that 6 i need not be small compared to unity.
For t - 0, and mi-*
Pc3L/
- It
'small vibrations, with Eg -+ 0, then g
h should be noted that the quadratic terms in Equation (13) o not behave as velocity squared damping terms, but appear to behave more like non-linear the unig rectional motion di This is easily seen by considering springs.
It is clear, there, that the term y N[ changes sign
= 0.
case with Wt only when E2 changes sign; that is, no net energy is dissipated duritg a
. complete harmonic motion cycle.
- 10. The usual small motion analysia Nhdas only the virtual mass In effects; the effect of fluid Triction can be shown to be negligible.
our case, where large motions, relative to the nominal gap width, h, are admissible, we need to compute additional contributions to Equation (13) which account for frictional forces and the turning losses.
The following approximate analysis is used to develop the additional terms necessary to include dissipative effects.
- 11. If the balance between pressure and frictional shear forces along any portion of the duct is considered, Reference 5 shows that the straight governing relation for the fluid pressure change due to shear stress is U
& =
f*p Ui I
3 P 3S h
h (16),
=-2 g
y y
where t' is the friction loss factor defined by the relation t* P U
Tfi " 2 i
i Equation (10), written for each portion of the duct, can be integrated 12.
to ' yield the pressuee distribution in each region. Note, however, tha't for arbitrary 2-D motions, proper attention must be focussed on the local fluid flow direction in cach duct in order to ascertain the location of points'or flow reversal. - The arbitrary time functions. arising from the integration of Equation. (16) arc determined by applying corner pressure matching conditions. _ From Figure 3, for 1, j combinations 1,2 : 2,3: 3,4 : 4,1,
empirical cquaticns zor pressure loss due to cbrupt turnin6 er the ritu may be urittcn C88 P.;
Pi
=-
V V
- 27, = Ug+U$
p-Where K is a loss coefficient.
- 13. - rce a general-2-D motion (51, $2 (f 0), the paure in each region is 1
fully determined by using Equations
- 16) and (17); contributions to the generalized forces Qft(Equation (13)), due to frictional and turning
. losses, are calculated from the expressions, c/ 2 F
=L (p 9 ) dy g
3 1
. C/2 (18)
F
=L (P - P ) dx y
4 2
-c/
2 Note that the direct errect of shear forces is neglected in Equatien (18).
- 14. For the purpose or illustration, we restrict further damping computations to uni-directional motion, say W2 p 0.
It is clear that the plane x : 0 is a plane or symmetry for the fluid flow; for such a motion, the pressure distribution due to triction losses in regions 2, 1,
4 in Figure 2 are obtained from Equation 116) as p f* }
x 2
2 P I*'t) =
- A (t) ; x
>,0 2
2 3
2 3c h (1-c I 2
I19) 2-p f* u W
"2 Y
c 2
+ h i Y4 2 P y (y, t) =
4h p f*
11 x
~
2 2
P 4 (x, t) =
2 3
4
+ A (t) ; xto 3c h (1-c I 2
Using Equation (18) to compute F yields y
c/ 2 3
p p f*y C.L W2 "2 F = 2L (p4 - p2) dx =
y 96 J
'(20) o 1
1
+ (A -A ) C L 4
7 (1+c )
(1~"2)3 2
Using Equation (19), and Equation (7) in Equation (17 hat the corner.s x e/2
- Y
- I.c/2 yields:
- - _ _ _ _ = _ _.
]
2 0s 2
L (2-c 1 3 -3 p
2 2
2 1
2 f, 1 + _1 0
4 (1-c I 3Il-C I 2
2 s
'(21) r 2}.2 2
W W (2+c Ai-A4
=y 2
2 K
1
+ uf, 1+
8 4
2
@cy 3(1+g2}
p A -^4 2
- 15. Solving Equation (21) for and substituting in Equation (20) p yields the resulting velocity squared nerz-linear damping force as Fy=C 6-2) 2 2.
~
~
2 2
4 (22) where (1+3c I
I4-3C2 +C2 I 2
2 K
> + T-pp CL 4 uf 1+
23 22 e (g I" 4(1-8 3
Il-C I
2 4
2 2
s Dynamic Analysis
- 16. The simple dynamic =ocel shown in Figure 4 is new considert ; as a vehicle to obtain numerical results which illustrate the effects of the K,
K are introduced representing the fluid coupling.
Linear springs I
o i
elastic stiffness coupling the structure to ground, lumped masses H, Ho y
I representing the mass of the respective solid bodies, and non-linear elements F ', Fg which acc.maty s:nen :mpact between the bodies cecurs.
e 2
Then if Y(t) represents a known ground motion, and relative co-ordir.ates U, y are introduced by the relations U2*U+IiY2=v+Y (23) the equations of cotion for the system shoun in Figure 4 can be written as (My + W ) d - m2 Y
U -I ' + F4' - HI 2
V U-E I
2 l
(24) 0v+F'-F'-Hoi l
- m2 3 + UIO + m2)
-U-E 2
4 where the impact elements F ' (hg) = 0 if h4>0 4
(25)-
F ' (h ) = 0 if h2>0 2
2 the fluid mass m2 is given as 2
y pc 3
1+
m2" 2
2 3f1-c I
2 anc the fluid forces are 2
L 2
G=
' ' ('2) 2 2
(26) 2 2
~
(1-c I
2
j '~
1*/.
It is cicar that if tha motions U, v are assumed small, then it may be obtaincd by including only the cffect r.rgurd that conservative results arv.u O c' L, and nesiscting tha effcet of G.
2 of a constant fluid mass However, if the effect of Isr$e motions is incorporated with respect to t_ hen more realistic re=ults can be achieved, since it is nominal gap size, easily shown that increases in the fluid effects encompassed in G are larger than the increases in fluid mass.
To illustrate the fluid effects, equations (24) and (26) a[,e solved, 18.
for a given time history ground motion Tit), using a modification of the Typical geometry and time history computer code presented in Reference. 6.
material values are assigned which are representative of a fully loaded fuel rack containing 169 cells simulated by a two degree of freedom dynamic that such a simple model is only for illustrative model.
Note, however, purposes; the authors have developed a mere realistic rack and fuel assembly group model which uses thirty-tuo degrees of freedom to accurately simulate potential for rack deformation, impact, and sliding under a realistic 3-D seismic event.
For the purpose of illustrating the effects of confined fluid, the following input data is used:
Rach Mass Ho = 9368.8 Kg Fuel Assembly Mass (169 cells) My = 53586.1 Kg e s.1524 m. : h =,7.9375 mm. ; L = 53 34 m 4
Ky : 60590.9 N/m. ; Ko = 60590.9 x 10 g/m.
f's 0.025 ; K =.9 6
3 10 ggfm p = 1019.71 Eg/m3 x 169 cens.w.l.Tl23 19 To simulate the impact force on the cell walls, non-linear gap elements with stiffness to Ko are used.
These gap elements become active, when the h2 or h4 approach value.01 h.
The seismic acceleration Y(t) used in the simulation is,Y(t) = A sin D t where A =.5 g 10 Herts 0 (t 4
.2 see (26)
= 5 Hertz t
.2 see A =1.0 g A total event duration of 1.3 sec. is assumed.
The following five simulations are performed using the two degree of freedom model:
Remarks Case Vibration in air; no fluid mass or damping 1
Vibration in fluid; small deflection model - E i 2
0 when calculating m2 and fluid damping effcet Vibration in fluid; large deflection model used 3
r-fcr cocputing fluid masses offset - na fluid i.**
f damping i-4 Same as case 3 except fluid damping included same as case 4, except f', K reduced to.1% of 5
values used in case 4.
Discussion of Results of Simulation Runs
.20.
The following table summarizes the results obtained from the five simulations.
Figures 5 and 7 show typical time histories of the rack spring force.
The magnitude of the rack spring force range is a direct measure of the expected rack stress level at the rack base and the subsequent loads transmitted to the pool floor slab through the rack feet (which are not modelled in our simple simulation).
TABLE 1 Summary of Results - Max. Force Range Case Rack Spring Force Local Impact Force Fluid Damping Range (N x 10-6)
Range from Cap ForceRagge Springs (H x 10 6)
(N x 10~ )
1 2.678 12.005 2
2.228 7.1 99
.756 l
3 1 997, 10.555 4
1.535 0.
1.503 5
1 766 7.499 4'.852 Examination of the maximum force ranges shown in Table 1 produces the
, anticipated results; namely,,the inclusion of fluid damping coupled with large deflection effects significantly reduces the force range in the fuel rack. Comparison of the results of case 4 with case 2 shows a reduction of 315 in the rack force range by the inclusion of large deflection effects in the calculation of fluid mass and fluid damping. The impacts with the cell vall are eliminated, thus eliminating the need for calculation of local impact effcets on the rack cell wall.
The results of case 3 indicate that at least for 'this simulated seismic input, the inclusion of only large deflection effects in the fluid virtual mass and the complete neglect of
, fluid damping serves to reduce the force range in the rack.
n o local stress range in the rack cell wall is increased in this ca,se however.
The authors hesitate to draw any lasting conclusions from the case 3 results since a change in input seismic frequence content may very well reverse the conclusions inferred from this data.
--___-_----_---_._.--_____._..-_-_-n_
I 21.
The results obtained in cass 5 terit some furthsr einbreation.
Tha riduction in fluid damping to 11, of th2 valu:2 u:cd in caen 4 is an attenpt i
to simulate the po::ibic damping effect of unchannelled foci cc:cmblics.
It is clear that the damping and virtual ma:s effects from an unchannelled fuel a::cmbly should be substantially les since the confined fluid has more unobstructed area in which to flow as the fuel as:cmbly moves relative to the cell wall.
In addition, there are substantial differences in the flow-field which shculd be considered in any analysis of uncha'nnelled fuel.
Nevertheless, case 5 may give come indication of what might be expected if i
only unchannelled fuel as:emblic: are in the rack.
Table 1 shows that the f
rack force range certainly increases over the results obtained in case 4; the rack force level is still substantially less than the results obtained for case 2.
Local impact with the rack cell vall occurs during the event althougn the impact range is less than that of case 2.
A somewhat surprising re:ults, on first reading, is obtained for the maximum fluid damping force range. Since the damping coefficients have been reduced, one might expect that the damping force range should also be reduced. However, we recall that the damping force is of the form C(6.)Ol0, (27)
Fd
~
W is maximum when Ev 0, and goes to zero as the gap closes.
C (C )
achieves its largest value when the gap closes, and is relatively==all when E.~ 0.
Examination of the detailed nut:,erical output from the simulations show that the damping force exhibit.s a sharper and higher peak in case 5, compared to case 4, but the energy dissipation due to the fluid damping is higher in case 4.
The increased dissipation in case 4 precludes C (E ) from growing too large since the gap never becomes too small.
The effect of fluid damping cn the rack spring force range is measured by the dissipation level during the event, rather than the peak value of the fluid damping force.
Thus, the expected result that a decrease in the effective fluid friction coefficient results in increased rack force level is obtained.
22.
Examination of the detailed ti=e history of rack force level shows that in air the rack essentially vibrates at its natural frequency of 41 Hert: with amplitudes modified by the local impact forces.
Although not shown here in the r sults, during the 1 3 second, ti=e of the event, vibration in air results in a total of twenty-two impacts with the outer cell walls.
With the additional of fluid mass, the graphs show that the rack essentially vibrates at the forcing frequency of either 10 or 5 Hertz.
l The addition of fluid mass effects in cases 2-5 reduces the number of
)
impacts to a total of three during the time span of the event.
CONCLUSIONS 23 It has been demonstrated that in high density fuel racks containing
{
channelled fuel assemblies, large displacement effects coupled with the f
inclusion of fluid damping results in a significant decrease in rack force-range and possibly the complete elimination of local Ampacts 'setween rack cell and fuel. assembly.
It has also been shown f that an approximate
- analysis of the large displacement effect is easily i=plemented into a ti=e I
history lumped mass analysis.
In a 3-D motien cross coupling effects between the two on-plane motions will occur in the inertial terms due to I
l
l 33rgo dispinecments; although not carried out in detail herein, similar
,fross coupling in the fluid damping terms is expected.
24, Experimental work is currently planned to verify the analysis presented here.
Once the analysis has been matched with *experi=ent, for both channelled and unchannc11ed fuel, the accurate inclusion of. fluid damping effects should become an accepted feature of. the 3-D dynamic analysis of high density fuel storage racks.
REFERENCES 1.'
Sharp G R and Wenzel V A.
" Hydrodynamic Mass Matrix for a Multibodies System", Journal of Engineering for Industry, Trans. of the ASME, pp.
611-618, May 1974.
2.
Fritz R J.
"Effect of Liquids on the Dynamic Motions of Immersed Solids", Journal of Engineering for ' Industry, Trans.. of the ASME, February 1972, pp. 167-173.
3 Stokey W F and Scavuzzo R J.
" Normal Mode Solution of Fluid-Coupled Concentric Cylindrical Vessels", Trans. of the ASME, Vol.100, Journal of Pressure Vessel Technology, pp. 350-353, November 1978.
4.
Dong R G.
" Size Effect in Damping Caused by Water Submersion", AICE, Journal of the, Structural Division, May 1979, pp. 847-857 5.
Li V H and Lam S H.
Principles of Fluid Mechanics Addison Wesley, 1964, pp. 273-278.
6.
Levy S and Wilkinson J P D.
The C-.~ent Element Method in Dynamics, HoGraw Hill, 1976, Chapter 3.
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Question 13 0-Margins over Allowable for the SFP Floor and Walls Load Combination Location Mar;;in Dead + Live + OBE Seismic Pool Floor 1.01 North Wall 1.25 South Wall 1.67 East Wall 1.04 West Wall 1.08 Dead + Live + OBE Seismic + Operating Thermal Pool Floor 1.69 North Wall 1.06 South Wall 1.39 East Wall 1.49 West Wall 1.61 Dead + Live + SSE Seismic + Accident Thermal Pool Floor 1.67 North Wall 1.06 South Wall 1.41 East Wall 1.12 West Wall 1.01 I"Edu"Nouimumcuammim ooc
ENCLOSURE 4 RESPONSES TO RAI GROUP 3 QUESTIONS TilERMAL IIYDRAULIC AND LIFTING
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Question 1.
Section 5.0," Thermal-11ydraulic Considerations," states that the spent fuel pool (SFP) bulk pool temperature is 170 F. Section 8.3.d," Temp"erature Loads," states that the SFP is designed for a normal operating temperature of 170 F. Ilowever, the FSAR discusses a peak temperature 171 F for a scenario that reflects the normal refueling practice at Vogtle.
a) '
Explain the difference between the bulk pool temperature, the normal operating temperature, and the peak temperature.
' b)
The 170 F limit exceeds the American Concrete Institute (ACI) Standard 379, which states, in part, "for normal operation or any other long term period, the
[SFP] temperature shall not exceed 150 degrees." Provide an evaluation on how the SFP temperature would be above 150*F, and the justification for why the long'F limit is acceptable to meet the intent of ACI 3797 Also, explain 170 of these elevated temperatures on the SFP system, structures, and components?
Response to Question 1(a)
The " bulk" temperature is the global temperature of the pool. It is derived from the heat transfer calculations and is the temperature result if the spatial temperature distribution in the pool is ignored. The " peak" temperature (FSAR Table 9.1.3-4) refers to the peak value of the bulk temperature as the pool temperature goes through a transient (increase and subsequent decay) after discharging fuel into the pool. In the heat transfer analyses documented in Section 5.0 of the report, transient effects were excluded thus, the bulk temperature and the peak temperature are equivalent and represent a steady-state bulk pool temperature.
- For the structural considerations in Section 8.0 of the report, the " normal operating temperature" referred to in Section 8.3.d is the steady-state bulk pool temperature from
. Section 5.0 of the report.~
Response to Question 1(b)
For the structural considerations in Section 8.0 of the report, the bulk pool temperature of 170'F was defined as the " normal operation temperature." This is not to imply that the j'
Unit 1 SFP will be operated at 170 F over an extended period of time. Rather, this SFP l
temperature may be realized during cutage refueling operations and therefore was chosen as the limiting temperature for non-accident conditions in the Section 8 analysis. The l
pool temperatures for long term operation remain as described in the FS AR, below the l
- 150*F requirement of ACI 349.
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Question 2 Section 5.3," Discharge / Cooling Alignment Scenario," states that the bulk pool temperature must be maintained below a limit of 170 F for the discharge / cooling alignment scenario. Is this scenario equivalent to the maximum normal refueling case l
that is documented in the Vogtle FSAR? Does the scenario assume that the entire core is unloaded into the Unit 1 SFP at 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> after the Unit 1 reactor shutdown?
Response to Question 2 The scenario described in the FSAR assumed that the first assembly was moved to the spent fuel pool no sooner than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after reactor shutdown and that the entire core is instantly unloaded into the pool 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> after the reactor shutdown. The scenario describes what is the normr.1 practice at VEGP, i.e., the entire core is un!oaded for each refueling outage. In the FSAR, this scenario is called the " maximum normal refueling case".
The discussion in Section 5.3," Discharge / Cooling Alignment Scenario",is not intended to describe a specific omoading scheme. This section, in conjunction with Section 5.8.1,
" Decay Heat Load Limits", is intended to describe that in order to maintain a bulk pool temperature not to exceed 170 F, the total heat load in the pool must not exceed the value 6
of S t.87x10 BTU /hr as shown in Table 5.8.1.
Question 3 The maximum emergency core unloading case for Unit 2 is described in the Vogtle FSAR. This case assumes that the entire core is unloaded into the SFP at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after the emergency shutdown of the reactor. It also assumes that 84 assemblies from the most recent refueling, with a decay time of 36 days, and 1,821 assemblies from prior refuelings are present in the pool. For this case, the SFP temperature is maintained below I82 F.
Provide the analysis of the SFP temperature for the maximum emergency core unloading case on Unit 1.
Based on NRC guidance, the 36-day assumption is related to a 30-day refueling outage.
Is a 30-day refueling outage assumption applicable for Vogtle? For example, if a refueling outage is performed in 20 days, then the decay heat load for the refueling load should be based on 26 days rather than 36 days. What affect does this assumption have on the decay heat load limit for the maximum emergency core unloading case?
R_esponse to Question 3 A new analysis for the maximum emergency core unloading case was not performed. The Unit 2 pool has a capacity of 2098 assemblies. The heat load and temperature analysis for the Unit 2 pool bounds the Unit 1 pool with a capacity of 1476 assemblies. The Unit 1 E4-2
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and Unit 2 spent fuel pool cooling systems are identical so the bounding analysis applies to both units as it does in the current FSAR.
An engineering evaluation was performed for one refueling load after 26 days decay versus 36 days decay, and a full core (with equilibrium decay heat load) emergency offload after 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> decay. The difference in heat loads was about 2x10' BTU / hour.
This increase in heat load would result in a small increase in the pool temperature in the l
case of 26 days decay versus 36 days decay. So long as the decay heat load is managed so 6
l as not to exceed 58.13x10 BTU /hr (FSAR Tables 9.1.3-1 and 9.1.3-4), the pool temperature will remain within 182 F with a single train of cooling. This is conservative relative to SRP 9.1.3 which only requires that bulk pool boiling not occur with two trains ofcooling Question 4 Section 5.4," Decay Heat Load Limit," states that a conservatism in the decay heat load limit calculations is to base the thermal performance of the SFP Cooling and Purification System heat exchanger on the design maximum fouling level. Explain the basis for this conservatism, and how it is factored into the analysis.
Response to Question 4 The term fouling refers to mineralogical and biological deposits that form on heat exchanger heat transfer surfaces. The presence of these deposits reduces the heat transfer efficiency of the heat exchanger. The heat transfer efficiency at the design basis maximum fouling therefore represents the minimum heat removal capacity of a heat exchanger. The in-service thermal performance of a heat exchanger will be better than the assumed maximum fouling level performance. Therefore, for performing thermal-hydraulic evaluations, the assumption of heat exchanger performance at the design basis maximum fouling is conservative.
The decay heat load limit computed for the Vogtle Unit I spent fuel pool monotonically increases with increasing heat rejection capability of the spent fuel pool cooling heat exchangers. Therefore, maximizing the fouling level will subsequently minimize the heat exchanger performance and the corresponding decay heat load limit. For this reason, the heat exchanger temperature effectiveness of the Vogtle Unit I spent fuel pool coolers is based on the terminal temperatures at the maximum fouling level.
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Question 5 Is there uncertainty factored into the decay heat load limit that is discussed in Section 5.47 If so, then what is the uncertainty factor?
Response to Question 5 When using median values in a calculation, the uncertainties inherent in the data should be accounted for. If conservatively bounding values are used, however, then the application of additional uncertainty factors is not mathematically necessary.
The Vogtle Unit 1_ decay heat load limit calculation is based on a lower bound heat removal capacity of the spent fuel pool cooling heat exchangers, and several other conservative heat rejection mechanism ' assumptions:
(i)
Neglect conduction heat transfer through pool walls and slab (ii)
Neglect forced convection cooling of pool surface by HVAC system (iii)
Building ambient at design basis maximum temperature and 100% relative
. humidity (iv)
Neglect thermal inertia of pool water, fuel racks and stored fuel assemblies The use of these bounding values yields the calculation of a lower bound decay heat load limit. Any uncertainties,in the cycle-specific decay heat calculations that will be performed prior to each discharge from the reactor to the spent fuel pool will be accounted for in those calculations.
Question 6 In Section 5.5, what is the boiloff rate of the Unit 1 SFP7 Explain how the boiloff rate is derived? What is the capacity of the SFP makeup system and the rate at which it supplies the Unit 1 SFP?
Response to Question 6 The boil off rate of the Vogtle Unit I spnt fuel pool, at the decay heat load limit, is 5.347x10' lb/hr (approximately 14.9 ft / min er 111.5 gpm). This value is calculated by dividing the decay heat generation rate (assumed equal to the decay heat load limit) by
.the latent heat of vaporization of water at 212*F.
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The primary source of makeup water for the spent fuel pool is the Refueling Water L
Storage Tank (RWST). The RWST has a total capacity of 715,000 gallons and can be provided to the pool at a rate of 200 gpm.
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Question 7 Section 5.6 states that the bulk pool temperature is artificially reduced to conservatively maximize the fluid viscosity. Explain what is meant by " artificially reduced," and how it will maximize the head losses for water flowing through the fuel racks and fuel assemblies.
Response to Question 7 The friction factor for flow in the laminar flow regime in the fuel rack cells, a key component of hydraulic resistance to flow, is inversely proportional to the Reynolds 3
number. The Reynolds number is inversely proportional to the viscosity of the flowing fluid. Maximizing the viscosity will minimize the Reynold's number, thereby maximizing the friction factor and the corresponding hydraulic resistance. This results in 1
a reduced flow through the fuel rack cells and the calculation of an upper bound temperature rise in the cells.
The viscosity of water increases with decreasing temperature. In order to ensure that the water viscosity and the resultant hydraulic resistances are maximized, the temperature of the water entering the spent fuel pool is set lower than normal. This has the effect of reducing absolute temperatures throughout the pool, thereby maximizing the viscosity.
Because all local water and fuel cladding temperature calculations are based on bounding temperature differences, this global temperature reduction yields conservative results.
Question 8 Table 5.6.1," Data for Local Temperature Evaluation," provides a value of 60,000 mwd /MTU for the maximum average burnup. Ilowever, the Vogtle FSAR assumes a fuel burnup of 50,000 mwd /MTU. Why are these values different? Are there any other changes in the proposed submittal that have occurred that would affect the existing FSAR analyses?
Response to Question 8 An average burnup of 60,000 MWD /MTU was used in the local pool temperature analysis to maximize the decay heat and hence, the calculated local temperature for the Unit I racks. In the corresponding analysis for the Unit 2 racks, an average burnup of 50,000 MWD /MTU was used. If, at a later time, cycle designs change such that the average burnup will exceed 50,000 MWD /MTU, the Unit I local temperature analyses will not have to be revised. The use of an average bumup of 60,000 MWD /MTU for this analysis does not affect other burnups referenced in the FSAR.
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Question 9
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In Section 5.8.1,"Results of Decay lleat Load Limit Evaluation," provide more detail on how the maximum decay heat load limit of 51.87 x 10' Btu /hr was derived. Explain how the operational limits (i.e., in-core hold time) and the administrative controls will ensure l
that the heat load limit will not be exceeded for each refueling outage. Will this heat load l
be reflected in the FSAR since it is lower than the heat loads currently documented in the l
FSAR7 1
Response to Question 9 This question will also be further discussed in the response to Question 11.
The heat load limit of 51.87x10' BTU /hr is not predicated on any particular fuel management / discharge scheme. Rather, it is the stcady-state heat load that will maintain the pool bulk temperature at 170 F with a single train of cooling in operation. The limit of interest here is the pool temperature. Prior to offloading a core to the pool, a heat load evaluation will be performed to assess the impact of the offload on the total pool heat load. By ensuring that the total heat load from prior offloads and the core to be discharged does not exceed 51.87x10' BTU /hr, the pool temperature will not exceed 170 F with a single train of cooling. The evaluation will determine the necessary in-core hold time to ensure this limit will be met. Procedures will be used to administratively control heat load management of the pools..The FSAR will be updated to include the heat load that will ensure the temperature limit of 170 F will not be exceeded and will also include the requirement to perform a heat load evaluation prior to transferring irradiated fuel to either pool.
Question 10 in Section 5.8.3.," Local Water and Fuel
.ang Temperature," explain how the bounding peak cladding temperature 228"F was derived.
Response to Question 10 The bounding peak cladding temperature is obtained by adding a number of conservative temperature difference values to the maximum bulk temperature limit (170 F). This procedure is performed as:
a.
The difference between the peak local water temperature and the bulk water temperature (ATw) is extracted from the results of the computational fluid i
dynamics (CFD) evaluation. The CFD evaluation is performed in a conservative fashion that yields a conservative value for this parameter.
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b.
The bounding peak local water temperature (T ) is determined by adding the i
bounding bulk-to-local water temperature difference to the maximum bulk temperature limit (T ).
3 c.
A bounding local water to fuel cladding temperature difference (ATu) is calculated using principles oflaminar flow heat transfer.
d.
The bounding peak fuel cladding temperature (Te) is determined by adding the local-to-clad temperature difference to the bounding local water temperature (T ).
i Parameter Value T
170.0 F 3
+ ATw 15.1 F
=T, 185.1 F i
+ ATu 43.1 F
= Te 228.2 F Question 1i What are the major differences between the decay heat load limit analysis that was performed on Unit 2, and the current decay bat lead limit analysis that was performed on Unit 17 Response to Question 11 The discussion of the refueling cases in the current FSAR Section 9.1.3 is applicable to both Unit I and Unit 2 spent fuel pools. For the refueling cases discussed,it can be seen that the control of pool temperature is dependent on heat load management rather than on any particular fuel management / discharge scheme. For example, consider the " maximum normal refueling case"in the current FSAR, i.e., a full-core offload which is the normal practice at VEOP. To maintain the pool temperature within the limit of 171.1 F with one 6
train of cooling in operation, the total heat load must not exceed 54.1x10 BTU /hr (FSAR Tables 9.1.3-1 and 9.1.3-4). For the example used in the FSAR, if the heat load from the full core is subtracted from the total allowable heat load, then the maximum 6
allowable heat load from prior offloads is limited to ll.98x10 BTU /hr. If the heat load in the pool is greater than this value, then prior to offloading a full core, a heat load evaluation has to be perfenned to determine the additional core decay time required so that the total heat load will not exceed 54.1 x10' BTU /hr to control pool temperature to within the limit of 171.1 "F.
Previously, a projected fuel discharge scheme was used to simulate the loading of the Unit 2 pool. Based on this scheme, it was shown that with the pool essentially at capacity, the addition of a full-core offload would not result in the bulk pool temperature exceeding 171.1 "F. Since fuel management, i.e., numbers of assemblies and bumups, E4-7 l
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Wha' is ofimportance is to manage the total heat load and therefore control the pool temperature to within the specified limit.
The revised heat load analysis concentrated on determining the maximum total heat load to control the pool temperature to within a limit of 170 F. This temperature was chosen since it is within what has been previously licensed for VEGP. The heat load was not predicated on any particular fuel management / discharge scheme. Since the spent fuel pool cooling systems for the Unit 1 and Unit 2 pools are identical, the same temperature limit and corresponding heat load will be applied to both pools. The limits discussed in FSAR Section 9.1.3.1 will be revised based on the revised analysis.
Again, it should be emphasized that the controlling parameter is the pool temperature.
The pool temperature will be controlled to within its limit by managing the heat load in the pools. As discussed in the response to Question 9, the pool heat loads will be managed by administrative controls. These controls will be placed in the applicable procedures. As discussed in the response to Question 9, the FSAR will be updated to include the heat load that will ensure the temperature limit of 170 F will not be exceeded as well as the requin ment to perform a heat load evaluation prior to transferring irradiated fuel to either pool. The discussion in FSAR Section 9.1.3.1 for the maximum emergency unloading case will also be revised to clarify that a similar heat load evaluation will be required to ensure that the temperature limit in the FSAR is not exceeded.
Question 12 Demonstrate how the temporary gantry crane (rated for 20 tons) will be erected on the existing spent fuel bridge rails; discuss how the crue will be used to handle rack removals and installations.
Response to Question 12 See response to Question 13.
Question 13 Briefly discuss at what point the Fuel llandling Bridge crane will set down the racks and the temporary gantry crane will be used to lif Jae racks as the racks are transferred to and from the fuel building.
Response to Question 12 and 13 The temporary gantry crane will be erected in the truck bay area. Each vertical leg will be uprighted onto prefabricated rail supports with support scaffolding. The top portion of the crane will then be installed followed by the hoist. The temporary gantry crane will E4-8
then be lifted using the main overhead crane and transported to the fuel handling crane j
rails. The temporary crane will be lowered onto the rails, seismic clips will be installed, and the temporary gantry crane will be connected to the fuel handling crane. Mechanical stops will be installed on the crane rails to prevent any crane travel over the Unit 2 spent fuel pool.
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The temporary gantry crane will transport the existing racks from the pool and install the l
j new racks into the pool as follows:
l North / South Direction - Driven by lioist on Temporary Crane East / West Direction - Driven by Fuel Handling Crane The existing racks will be removed from the pool using the temporary crane, transported to a transfer pad installed in the new fuel storage area located adjacent to the Northwest corner of the pool, disconnected from the temporary crane, and transported to the truck bay with the main overhead crane. This procedure will be reversed for installation of new racks.
Question 14 Explain why the existing fuel racks will be lifted just above the pool floor and held at that elevation for a length of time before making any further rack movements; also discuss how much time is being considered for suspending the racks above the pool floor.
Response to Question 14 The proposed plan for the testing of the lifting components within the scope of the rack removal and reinstallation, includes the load testing of a temporary gantry crane, spreader beams, and other lifting devices. Each of these devices will be load tested separately prior to their initial use within the Unit I spent fuel pool.
Per the guidance of the ANSI B30.2, Para. 2-2.2.2, the intent of the code is to load up and down the runway (crane rails) for the full range of service load lif ting. The code requires that the load test be performed at 125% of the service load requirements. To meet the intent of the code and minimize the potential damage to the pool liner, we are planning to lift the existing racks (which are the heaviest load) approximately 6 inches off the pool floor and hold them for 10 minutes (+/-) to provide added assurance that the installed load lifting " system" is functioning properly prior to any movement. This lift height will minimized the effects from an accidental load drop during this period.
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Question 15 Discuss why the refueling accident analysis involving a dropped fuel assembly weighing 2300 lbs. and the potential consequences is used instead of using a bounding analysis involving the dropping of the existing racks weighing 33,000 lbs during removal of those racks.
Response to Question 15 The refueling accident analysis was per' formed pursuant to the provisions in the USNRC "OT Position Paper"(ca.1978). This analysis is focussed to quantifying the structural damage sustained by the fuel rack due to a postulated fuel assembly drop during fuel handling operation. Safety considerations collateral to this analysis pertain to insuring that (i) the plastic deformation at the top of the rack structure does not lead to local boiling in a storage cell due to constriction of the thermosiphon flow and (ii) that the I
active neutron absorber region in the rack is not subject to plastic deformation (which will produce a change in the system reactivity).
The drop of a fuel rack during the rack installation effort need not be postulated in this 1
project because:
(i) There will be no fuel in me pool at the time of rack movements, and therefore there is no credible scenario of a radiation, criticality, or thermal-hydraulic event in the pool from a rack drop.
(ii) Careful attention has been paid to insure that there are no safety related or important-I to-safety equipment in the locus of the heavy load travel path.
1 (iii) Defense-in-depth measures set forth in NUREG-0612 to prevent rack handling accidents will be implemented during the rack installation work through controlled procedures.
Question 16 Briefly discuss the evaluation method used to determine whether the lifting devices and associated lifting components satisfy criteria I-IV in Section 5.1 of NUREG 0612.
Response to Question 16 l
The lifts proposed during the rack installation and removal do not involve any lifts over or close to spent fuel. Therefore, Criteria 1 addressing a radioactive releases outside the limits of 10 CFR 100 and Criteria Il addressing a change of fuel configuration such hat l
"k" effective is larger than 0.95 does not apply. Since the fuel in Unit 1 pool will be removed prior to any rack replacement activity in the Unit 1 SFP, there is no potential for damage to the SFP which will uncover fuel as described in Criteria III. Finally, the lift
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f areas, load capacities, and load paths have been analyzed for a postulated heavy load drop l
as required by Criteria IV. The damage caused by a potential load drop is minimized by the lift heights required to be maintained throughout the buildings. Maximum heights were calculated to prevent penetration and spalling of concrete during a potential load drop. In situations where the load heights must be exceeded, redundant rigging and a single failure proof crane will be used.
Question 17 Although the spent fuel cask is not to be carried over the SFP, briefly discuss the load path with regards to the SFP.
Response to Question 17 The cask handling bridge crane traverses the auxiliary building and a ponion of the fuel handling building. The cask handling crane's path is perpendicular to the path of the spent fuel bridge crane and is designed such that the cask crane cannot pass over the spent fuel pools. This precludes the movement of heavy loads (other than those associated with the spent fuel bridge crane) over the spent fuel pools. By means of built-in limit switches, the crane is restricted from traveling near or over the spent fuel pools when the main hoist is handling loads in excess of 15 tons, as shown in FSAR figure 9.1.5-1. This
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is discussed in detail in FSAR Sections 9.1.2.2, Facilities Description and 9.1.5.2.2, Spent Fuel Cask Bridge Crane.
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l Question 18 l
ANSI N14.6/1978-1993 defines "special lifting devices" as lifting devices for radioactive material containers for which the use of simple slings or chains, with or without spreader l
bars, is not appropriate because of special requirements, such as the need for remote engagement or for safety considerations. On page 3-3 of the submit'al, the licensee states that the lifling devices are not "special lifting devices." Explain why the lifting devices used in the proposed load handling operation is not needed for safety considerations.
Response to Question 18 During the proposed load handling operction, SNC has adopted several measures to minimize safety concerns from guidance provided by NUREG 0612.
The effects ofload drops have been evaluated for each unique area within the load paths of the rack replacement project.
In addition, procedures, drawings, and mechanical stops have been created to control the load handling process and to eliminate the potential for the existing or new racks to come close to spent fuel or exposed safety-related equipment.
Load paths distances relative to the Unit 2 Spent Fuel pool have been maximized.
Also, rail stops will be installed to prevent refueling machine / temporary gantry crane from being positioned too close to the Unit 2 pool.
Also, the use of redundant rigging will be utilized in those situations, where load drop heights exceed the allowable height requirements.
The ANSI N14.6 standard intent for "need for remote engagement or safety considerations" is applicable for those cases where manual engagement of rigging is not possible due to lack of accessibility to the lift interface points or is so unsafe from a life safety standpoint. This would be the case where the access to the lift points would be l
prohibitive due to high radiation fields and/or the proximity required to engage the lifting j
mechanisms would put personnel in a compromised safety situation. Neither of these conditions are applicable to this re-rack operation.
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I ENCLOSURE 5 l
RESPONSES TO ADDITIONAL QUESTIONS I
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Question Regarding Solid Waste In terms of our annual or lifetime projections for solid waste generation, how much solid waste will be generated as a result of the Unit I re-racking?
Response
Our current plans are to have the old racks removed intact from the site by a salvage l
company. After the usable material has been salvaged, the remainder will be volume reduced and disposed of at Barnwell against the Vogtle permit issued by South Carolina.
In the worst case if the racks have to be disposed ofintact, i.e., no material salvaged and no volume reduction, the maximum volume to be disposed of will be approximately 44%
of the projected solid waste volume for 1998.
Question on Codes and Standards Discuss the potential differences between the codes and standards used for the design and fabrication of the MYAPC racks and the codes and standards required at VEGP.
Response
The following is a discussion on potential differences between the codes and standard used for the design and fabrication of the Maine Yankee racks and the codes and standards required at VEGP. In the SER described above, section 2.4.2 " Applicable Codes, Standards and Specification", states that the design, fabrication, and quality assurance standards for the Maine Yankee spent fuel racks were compared with the NRC's "OT Position for Review and Acceptance of Spent Fuel Pool Storage and Handling Applications" dated April 1978, including a revision dated January 1979 and were found to be acceptable. Our licensing report chapter 1.0 " Introduction" pg.1-1 also states that the design and analyses performed meet all Vogtle applicable codes and standards, in particular the above referenced "OT Position" paper.
The "OT Position" paper allows for the design, fabrication and installation of spent fuel racks of stainless steel material using AISC specification or Subsection NF requirements of Section III of the ASME B&PV for Class 3 component supports. The Maine Yankee racks used AISC requirements and the Vogtle racks to date have used the ASME requirements. The design of the Maine Yankee racks have been analyzed to the ASME requirements applicable to Vogtle and found to be acceptable. We therefore conclude that the design and fabrication of these racks are in compliance with the codes and l
standards applicable to Vogtle and previously approved by the NRC.
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Que-tion on Venting of Boral and Performance of the MYAPC Racks Are the boral enclosures vented? What has been the performance of the racks with respect L!ging or swelling of the cells.
Response
Documentation from Maine Yankee indicates that the racks have two 1/4" vent holes at the top of each cell at opposite comers and one 1/4" drain hole at the bottom of each cell.
Per conversations with Maine Yankee and a review of documentation received, Maine Yankee experienced some bulging of cells in five racks of a vintage prior to the ones shipped to VEGP. The affected cells were later " drilled and vented and resolved the bulging problem". Because of this, a commitment was made to perform a surveillance of the racks. Maine Yankee implemented a Spent Fuel Rack Sun eillance procedure once per cycle,just prior to each refueling outage. This procedure required a statistical sampling of the cells, and was in place during the entire time that our racks were used at Maine Yankee. The results of the last surveillance performed on these racks indicated that 39 cells were dragged tested and 11 cells were visually inspected with none showing signs of swelling or bulging. Results of the previous two surveillance also showed no signs of swelling or bulging. During the process of removing the racks to be shipped to VEGP, a considerable amount of fuel shuffling had to be performed and no difficulty was noted.
After the NRC issued their SER, Maine Yankee decided to flood the cavity where the boral was located in order to increase the flux trap area. This was done by drilling two 1/4" holes at the top and one 1/4" hole at the bottom of each cell. By installing these holes they also were able to prevent the potential build of gases in the cavity which could lead to swelling or bulging of the cell walls.
Prior to placing into senice for fuel storage, the racks will have had each cell drag tested for acceptability. The VEGP fuel handling machine also has a load cell with a trip function to ensure that an assembly is not lifled with excessive drag force.
Question Regarding Stainless Steel in Racks Section 2.5.1 of the MYAPC SER dated June 16,1982, states that the adjusting bolts of the rack feet are made from Type 17-4 Pli stainless steel. Section 3.2 of the VEGP Modincation Report references precipitation hardened A564-630 stainless steel. What is the difference?
Response
Type 17-4 Pli stainless steel heat treated at 1100 F and ASTM A564-630 are the same material. The 1990 Annual Book of ASTM Standards, Section 1, Vol. 01.05, Designation A564/A 564M-89, describes several material standards for Type 630. Type 17-4 PH stainless steel heat treated at 1100 F is one of the descriptions. The vendor drawing for the adjusting bolt of the rack feet uses the ASTM designation.
Clari0 cation to Response to Enclosure 4 Question 1 (b)
As requested in Question 1(b), provide an evaluation on how long the SFP temperature would be above 150 F.
Response
In the license amendment request for Power Uprating of the VEGP units (letter ELV-03375, February 28,1992, Docket Nos. 50-424 and 50-425, TAC-M82724 and 82725),
Section IV.D of the Balance-of-Plant (BOP) Licensing Report discussed the results of the evaluation of the spent fuel pool cooling system. This section of the report is the basis for the heat loads and temperatures for the three refueling cases discussed in Section 9.1.3 of the current FSAR.
For the maximum refueling case, i.e., a full-core ofHoad 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> after shutdown, after the temperature peaks at 171.1 "F, the temperature decays to below 150 F approximately 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> after discharging the entire core to the pool. Since a refueling outage occurs once per 18 months for a given unit, this is equivalent to about 3% of the total cycle time.
Typically, over half the fuel just of00aded will be reloaded into the core well before 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> so the actual time for the temperature fall to below 150 F will be considerably less than 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br />.
6 The temperature of 171.1 F and the corresponding heat load of 54.1x10 BTU /hr in the previous analyses referenced above are suf6ciently similar to the temperature of 170 F 6
and the corresponding heat load of 51.87x10 BTU /hr in the Unit I re-rack analyses.
Therefore, it can be concluded that the time referenced above is applicable to the new analyses for estimating the maximum time the temperature would be above 150 F.
The concrete walls and floor are several feet thick with a temperature gradient across them. At most, only a few inches of concrete will experience temperatures above 150 *F for short durations as discussed above. For the m.'s term durations between refuelings, the bulk pool water temperature, and hence the concrete temperature, will remain below 150 F.
Clarification to Response to Enclosure 4 Question 3 Explain why a new analysis was not performed for the emergency core unloading case.
Response
In this case, SRP 9.1.3 requires that bulk pool boiling not occur. VEGP previously licensed a pool temperature of 182 F for the emergency unloading case with a 6
corresponding heat load of 58.13x10 BTU /hr. The discussion in the FSAR will be revised to reflect that the temperature will be limited to 182 "F by controling the heat load and that a heat load evaluation will be performed prior to an ernergency unloading. This is discussed in the response to Question 11. The analysis previously performed for Unit 2 currently bounds Unit 1 and will continue to do so.
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.- Questions About Rack Design
- 1. What are the pool dimensions?
' 2. Who designed and fabricated the 26 racks for Maine Yankee?
- 3. Are the 26 racks on site now?
- 4. Provide the facters of safety from the spent fuel pool analysis (need only one table that shows the numbers (factors of safety) for the question marks shown below):
Shear Moment Axial - Moment & Axial Wall.. (North) 7 7
7-
?
(South) 7 7
7 7
(East)
?
?
?
?
' West) 7 7
7 7-Slab (i: orth) 7 7
7 7
(South)
?-
7
?
7 (East) 7 7.
7 7
(West)
?
?
?
?
Responses
- 1. Pool dimensions: 50 feet (L) x 34 feet (W) x 40 feet (D)
- 2. The racks were designed and fabricated by par Systems.
- 3. The 26 racks are on site at VEGP.
- 4. Requested data are in table below:
Factor of Safety Maximum Maximum Factor of Safety Shear Moment' Axial
- Moment & Axial (in-kips /ft)
(kips /ft)
Wall (North) 1.92 24,955 1,649 1.06 (South) 1.66 13,491 723 1.39
- (East) 4.96 13,491 723 1.04 (West) 1.36 9,062 566 1.01 Slab" 1.05 12,172 681 1.01
- Per discussion with NRC, the maximum moment and the maximum axial are provided rather than factor of safeties. Note that the maximum moment and the maximum axial force do not occur at the same location in the pool structure.
" The slab is not divided into North, South, East, and West areas so only one set of values is given for the entire slab.
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_ _ _ _ - _ - _ _ _ - _ _ _ - _ _ = _ _ _ _ _ _ _
Clarification to Question 2 of Enclosure 2 Provide additional details regarding the electronic dosimetry to be used during the reracking.
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Response
l Below is a description of the proposed monitoring instrumentation to be used to monitor the divers during the reracking operations.
Each diver will be monitored with multiple teledosimetry devices. Each teledosimetry device consists of an electronic direct reading dosimeter (Merlin Gerin DMC-100, utilizing a silicoa detector) connected to a transmitter by a phone cable. The transmitters will be clustered together on the back of the diver. The dosimeters will be located on the various body locations to be monitored. An underwater antenna (specifically designed for underwater use) will be connected directly to the back of the diver's suit in close proximity to the transmitters to maximize data reception. The dose and dose rate data from the dosimeters will be transmitted by the antenna directly to a radio receiver on the floor of the spent fuel pool area. The data will then be routed from the receiver to a computer and will be graphically displayed on the screen using a telen.etry software program. The technician monitoring the display will receive audible and visual feedback from the dosimeters whenever:
- 1) a preset dose or dose rate alarm occurs,
- 2) transmitter battery power is low, or
- 3) transmission with a dosimeter is lost.
In the event of a transmitter failure (due to loss of power or connection with the dosimeter), the dosimeter is powered with a separate battery which allows the dosimeter
~
l to continue to acquire the correct dose.
Underwater surveys will be conducted with the Merlin Gerin AMP-100 dose rate instrument. This instrument utilizes a GM detector and underwater cable. The maximum range of detection is 1000 rem /hr.
It should also be noted that there will be no irradiated fuel or other sources of high radiation in the Unit I spent fuel pool during the re-racking operation.
Clarification of Section 9.3 of the Modification Report Clarify the last sentence on Page 9-1 of the modification report.
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Response
The reference to Section 4 of this report is erroneous. The adequacy of the shielding around the pool in terms of dose mtes has been addressed in the response to Question 1 in
' of this letter.
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