ML20217K439
| ML20217K439 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 10/18/1999 |
| From: | Hartz L VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20217K441 | List: |
| References | |
| REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR 99-437, NUDOCS 9910260148 | |
| Download: ML20217K439 (32) | |
Text
1 VIRGINI A Et,ECTltlC AND Powen CoswAN v l
Ricustosn, VnWilNI A 23261 October 18, 1999 i
i i
U.S. Nuclear Regulatory Commission Serial No.99-437
)
Attention: Document Control Desk NL&OS/ETS R0 Washington, D.C. 20555 Docket Nos.
50-338/339 License Nos. NPF-4/7 j
' Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY j
NORTH ANNA POWER STATION UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION i
SUMMARY
REPORT ON USl A-46 PROGRAM in an April 1,1999 letter (Serial No.99-027), Virginia Electric and Power Company t
provided additional information to support the USl A-46 program submittal for North Anna Power Station Units 1 and 2. During recent telephone conference calls with the NRC to discuss the evaluation methodology for certain outliers, the staff requested j
additional information to cornplete their review of our A-46 program. The enclosure to l
I this letter provides the additional information for the following issues:
Justification for the use of " Method A" of GIP-2 o
Justification for spatialinteraction of cable trays Calculations for Condensate Storage Tanks (CST) and Refueling Water Storage Tanks (RWST)
Since our last telephone conference call on October 7,1999, we have located additional information to support the structural adequacy of the CST. The CST is completely enclosed by a reinforced concrete missile barrier. Therefore, the walkdown team could not visually verify the anchorage of the tank. The current A46 analysis of the CST treated the tank as unanchored and uses methodology based on GIP--2 and
' Brookhaven National Laboratory (BNL) Report 52361. Additional details of this analysis and margin of safety for tank buckling are provided in the enclosure, including the relevant section of BNL Report 52361. In ordor to further verify the seismic integrity of the tank, plant design drawings were reviewed in detail. This review indicates that the tank is supported at the top and is completely encased in a two feet thick reinforced concrete barrier.. The top support and the concrete shiciding were taken into consideration in the original seismic design of the tank, but the current analysis for.A-46
)
did not take any credit for the existence of the top support. The top support will prevent any overturning or sliding of the tank during a Safe Shutdown Earthquake. Additional
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discussion of the design / construction and justification of the seismic adequacy of the I
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9910260148 991018 PDR.ADOCK 05000338
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-CST, including drawings and skett,c. of the CST, are included in the enclosure to this j
letter.
l There are no new commitments made in this' letter, if additional information is needed to complete your review of the items discussed above, please contact us.
)
f Very truly yours,.
I
$/
Leslie N. Hartz Vice President - Nuclear Engineering and Services Enclosure cc:-
U.S. Nuclear Regulatory Commission Region ll Atlanta Federal Center 61 Foreyth Street, SW 4
Suite 23T85 ~
Atlanta, Georgia ' 30303 Mr. M. J. Morgan NRC Senior Resident inspector i
North Anna Power St'ation
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Enclosure l
^ Response to Additional Requests from NRC Staff on August 9 1999 and September 1 1999 I
(Provided in Addition to Virginia Power's Response on April 1,1999 to RAls)
Unresolved Safety Issue (USI) A-46 October 14,1999 i
.,l--
Virginia Electrie and Power Company 1
i.
North Anna Units l'and 2 I
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.s,
Request on August 9,1999 during telephone conversation between Virginia Power and the
.NRC:
Provide additional technical justification on the use of Method A in Table 4-1 of GIP-2 for the components for which the in-structure response spectra exceed 1.5xBounding Spectrum
. (or the Reference Spectrum). Note that the amplification factors between the free-field ground response spectrum and the in-structure response spectre (ISRS) are more than 1.5, therefore, justify haw this restriction, as stated on p. 4-16 of GIP-2, is satisfied for North
- Anna Station.
~
Virgir.la Power Response:
Background:
Justification for the use of Method A was provided in our RAI response dated April 1,1999.
The components for which Method A of GIP.2 was used during the implementation of the USI A-46 program at North Anna Station and their locations were listed in our USI A-46 summary report submittal (Ref.1) and in Table 1 of our RAI response dated April 1,1999 (Ref. 2).
Method A was only utilized for line mounted equipment (valves, temperature sensors, etc.) and only for seven elevations located within four different structures. Table 1 of the April 1 RAI response provides the percentage exceedances in spectral peaks (and ZPA's) over the spectral accelerations (and ZPA's) of the Reference Spectrum, as requested by the 'NRC. Subsequently, on August 9,1999, the conservative, design in-structure response spectra at the seven locations where Method A was used were provided to the NRC staff (Ref. 3). Virginia Power also had the 1.5 x free-field grouno response spectrum plotted on each of these ISRS to facilitate the estimation of the structural amplifications.
1 Discussion:
Additional technical justification for the use of Method A for those situations at North Anna, where this method was used, is provided. This consists ofinformation regarding conservatism that exists in the amplifications in computed Design Basis Earthquake (DBE) ISRS at North j
Anna, as discussed below.
1
. Conservatism in the Calculated In-Structure Response Spectra at. North Anna:
i Table 1 provides a brief description of the construction of the structures where Method A was used. As is evident, the structures in which the " Method A" components are mounted
- are typical nuclear power plant structures. No unusual or plant-specific situations were identified which would cause the amplification factors for these buildings to be greater than j
those in typical nuclear power plant structures. All 4 structures are heavily reinforced
]
concrete shear wall structures. The Unit 1 Main Steam Valve House, the Service Building, i
l and the Containment Intemal Structure are all founded on rock. The Unit 2 Main Steam
. Valve House 'is founded on a structural fill / crushed stone material above compressive Saprofite and sound. rock. Brief descriptions of the building construction and foundation configuration are given below. More detailed descriptions are contained in the FSAR.
j l
i 2
l Y
L
TABLE 1 BUILDING -
SUMMARY
DESCRIPTION OF BUILDING CONSTRUCTION Main Steam.
MSVHs are heavily reinforced, concrete structures extending both above Valve House.
and below grade, located radially around the north side and adjacent to the j
(MSVH).
Reactor Containment structure. The size of each MSVH is approximately Units 1 & 2 45 feet by 32 feet. The MSVH Unit 1 is founded on the rock. MSVH Unit 2
)
is founded on structural fill / crushed stone above compressive Saprolite and I
sound rock with concrete back fill against the Containment.
Service The Service Building is a multistory structure on the south side of the
)
Building (SB)
Turbine Building. It is approximately 272 feet by 70 feet area below grade, j
In general, foundations consist of a structural mat and strip footings. The
)
Service Building is primarily constructed with heavily reinforced concrete shear walls. At higher elevations structural steel braced frames and 1
columns make up the load resistance mechanism.
{
Containment The containment external structure is a steel-lined, heavily reinforced Internal concrete structure with vertical cylindrical wall and hemispherical dome, J
Structures with access openings and penetrations. The internal structure is a heavily j
reinforced concrete structure with shear walls. The external and internal structures are supported on the same fiat base mat below grade. The containment structure is constructed inside with an open cut excavation in i
rock. The structure is rock-supported. The base of the foundation mat is I
located approximately 67 feet below finished ground level.
]
The maximum Amplification ratios between the Design Basis Peak Spectral Acceleration
]
(PSA) and the Ground Response Spectrum at North Anna for all the locations where Method A was used are provided in Table 2 below. As can be seen from the last column in the table, the maximum amplifications range from 7.35 to 2.2. These values are clearly higher than the 1.5 amplification factor identified with Method A, but they are very reasonable considering the conservatisms involved in the design basis analyses.
J 3
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I j
TABLE 2 Structures, Elevations and Amplification Ratios where Method A of GIP-2 is Used Amplification Direction Peak Ratio (East-Spectral (PSA/ Ground Equipment West or Acceleration Frequency Spectral Mark Site North-In the ISRS Range accel, at the Number (s)
Stru.ture Elevation Condition South)
(g)
(hz.)
same frequency) 1-MS-PCV-Main Steam 306' Rock E-W l.65 8.0 to 12.0 4.95 101A,101B, Valve House, N-S 2.45 5.2107.0 7.35 101C Unit I l-MS-TW Main Steam 285' Rock E-W l.13 8.0 to 12.0 3.40 101A,101B, Valve House, N-S 1.27 5.2 to 7.0 3.81 101C Unit I l-MS-TV-Main Steam 277' Rock E-W 0.93 8.0 to 1 '. o 2.80 111A,111B Valve House, N-S 0.88 5.2 to 7.0 2.64 Unit 1 2-FW-FCV-Service 286' Rock E-W l.17 5.0 to 7.0 2.34 2478,2479, Building E-W l.05 9.7 to 14.0 3.94 2488,2489, NS l.1 3.0 to 5.0 2.20 2498,2499 N-S 0.73 9.0 to 14.0 2.74 2-MS-TW Main Steam 285' Soil / Rock' E-W l.55 5.7 to 7.2 3J2 201A,201B, Valve ilouse, N-S 1.72 4.0 to 5.7 3.44 201C Unit 2 1-RC-LT-Reactor 253' Rock E-W l.24 4.5 to 6.2 3.72 1470 Containment.
Internal N-S 1.24 4.5 to 6.2 3.72 Structure 2-RC-TE-Reactor 256'-4" Rock E-W l.3 4.5 to 6.2 3.9 2410,2413, Containment, 2420,2423, Internal N-S 1.3 4.5 to 6.2 3.9 2430,2433 Structure
- Note: Compressible Saprolite is about 12' to 20' deep, with 10' structural fill. Deconvolution effects were considered.
The fact that the conservative ISRS have high amplifications at some buildings and elevations within about 40' of grade is due to the conservatism in the Design Basis ISRS for the North Anna Plant and the methods used in their development. As stated in Reference 2, the significant sources of conservatism included the following:
I,ocation ofInput Motion (variation from the free field input location)
Ground Response Spectrum Shape Ground Motion Incoherence Time History Simulation Existence of Narrow Peaks Structural Damping Frequency (Structure Modeling)
Non-Linear Behavior (e.g., concrete cracking)
Peak Broadening and Enveloping 4
L
The degree of conservatism involved in each of these parameters is specific to the building being' analyzed and to the floor level being considered. Reference 2 quantifies ranges of
-conservatisms for many of these parameters. The specific conservatism for each of these
-D parameters utilized'within the development of the North Anna ISRS cannot be accurately quantified using simple conventional calculation techniques since these parameters are inter-l related in a highly complex manner.
However, a comparison between conservative design spectra and. median-centered type
, spectra demonstrates the level of margin that exists for nuclear structures. This is based on
. information developed and compiled by EQE International, Inc. and shown in Attachment 1.
This attachment contains'a comparison evaluation of overall margins between median.
centered analysis and design basis analysis for five nuclear plant structures at four nuclear facilities similar in ' construction, building frequency and damping to those at North Anna Power Station. The reinforced concrete shear wall structures are two Auxiliary Buildings
- (AB), a Reactor Building (RB) interior structure, an RB exterior shell and a containment interior structure which are typical of those found at Nuclear Power Plants. The ratios of conservative design spectra to median-centered spectra are 2.53,5.3,3.3,2.3 and 5.4. The mean of these ratios is 3.77; This value can be used to estimate the appropriate, realistic amplification factor at North ~ Anna structures. Table 3 stipulates a modified (reduced) amplification level for each of the seven locations at North Anna where Method A was used.
' The reduction factor utilized was the 3.77 mean value from the study documented in. Tije resulting modified amplification factors range between a maximum of 1.95 to less than 1. These modified amplification values are much closer to a median-centered type value. Based on the conservatism in the development of design basis spectra
' discussed above, we have concluded that a realistic median centered assessmeut for the seven subject locations at North Anna would result in an amplification factor of"about" 1.5. It is
- noted that the NRC has accepted a similar assessment of margins for the R. E. Ginna Nuclear Plant (Reference 4).
5
TABLE 3 Estimated Amplifications for North Anna Design Basis Maximum Reduced Amplification Amplification Structure Elevation Direction Ratio Ratio
- Main Steam Valve 306' Ilouse, Unit 1 N-S 7.35 1.95 Main Steam Valve 285' Ilouse, Unit 1 N-S 3.81 1.01 Main Steam Valve E-W 2.80 0.74 277' Ilouse, Unit 1 N-S 2.64 0.70 E-W 2.34 0.62 E-W 3.94 1.05 Service Building 286'
~
N-S 2.20 0.58 N-S 2.74 0.73 Main Steam Valve 285' IIouse, Unit 2 N-S 3.44 0.91 Reactor Containment' 253' Internal Structure N-S 3.72 0.99 Reactor Containment' 256'-4" Internal Structure N-S 3.9 1.03 l
)
- Reflects the reduction by a factor of 3.77 which accounts for some conservatisms based on specific examples from the 5 nuclear structures researched by SQUG.
l 6
)
i Conclusion.
l The discussion above leads to the conclusion that there are significant inherent conservatisms
'in the methods utilized at North Anna to calculate design basis ISRS. These conservatisms
]
are difficult to specifically quantify without conducting some fairly costly analytical
{
exercises. However, the generic information in Attachment I has been assembled by EQE I
and SQUG and shows that the mean margin between median centered and design basis analysis for specific examples from five nuclear plant structures (we have described why we l
believe the North Anna structures to be similar) is about 3.77. This leads to the conclusion that the amplification factor for North Anna would be about 1.5.
l I
Based on this,;it is concluded that the intent of GIP-2 requirements and restrictions for the use of Method A are met for the associated equipment evaluated at the North Anna Plant.
j
References:
- 1. Submittal of USl A-46 summary report to NRC for North Anna Power Station Units 1 and 2, Virginia Power letter Serial Number 97-246, May 27,1997.
- 2. Response to Request for Additional Information on Summary Report on USI A-46, Virginia Power letter Serial Number 99-027, April 1,1999.
- 3. Fax Transmittal to NRC - In-Structure Design Response Spectra at seven locations of North Anna Power Station where Method A is used and their comparison with 1.5x ground spectrum,. August 9,1999.
- 4. NRC Plant Specific Safety Evaluation Report for USl A-46 Program Implementation at the R.E. Ginna Nuclear Power Plant (TAC No. M69449), June 17,1999.
7
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ATTACIIMENT I Comparison of Peak Spectral Responses between Design Basis to Median Centered Spectra 8
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taas Tansessakhwas Pacific Gas and Bactric Company
Figure 8: Diablo Canyon Ground Motion 3
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g steMe caeves Power Plant IMit Pacinc Gas and Hectric Campany tems Tona seismic Preeram
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L Request on heptember 1,1999 during telephone conversation between Virginia Power and 1
the NRC:
Clarify the spatial interaction calculation method for cable trays provided in your RAI response
' dated April 1,1999 and the example that calculated deflection using this method, also provided l
in the same RAl.
- Response-It is recognized that the multi tier cable tray displacement evaluation method provided in our earlier response to this RAI may not accurately represent the end displacement of the multi-tier l
. cable tray during a scismic event. The displacement of a multi layer cable tray could be slightly larger than that calculated using the above method and as shown in Example 10 which was submitted in the RAI response dated April 1,1999. However, during our USI A-46 walkdown of the cable tray system at North Anna, no spatial interaction concerns were noticed. Consequently, an accurate determination of the end displacement of the cable tray support was not considered significant.
In addition, Seismic Capability Engineers performed tug tests to assess the displacement of the cable trays. There were more than adequate clearances between the cable trays and the nearest commodities such as piping, equipment, components and structures.
Therefore cable tray spatial interaction was judged not to be a concern during the walkdown.
1 I
1 l
p 1
t I-c
)
NRC Requests for Additional Information (RAI) on Condensate Storage Tanks (CST) and Refueling Water Storage Tanks (RWST):
- a. Request on August 9,1999:
1 Data (preferably physical) to support two tank carthquake demand and capacity calculations (RAI # 8) particularly with regard to inclusion of water hold-down force, sloshing and resulting stresses in tanks stability and anchor bolt adequacy of the tanks.
- b. Request during telephone conversation on September 1,1999 between Virginia Power.and NRC:
)
.For the buckling (both elephant foot and diamond buckling) of the Condensate Storage Tank (CST), provide the margin of safety, Virginia Power Response:
Response to Request a.:
In our response to RAI # 8 on April 1,1999, the USI A-46 calculations for the two tanks, i.e., the Emergency Condensate Storage Tank (CST) and the Refueling Water Storage Tank (RWST) were provided to the NRC Staff. Both of these tanks were classified as outliers. In addition, the CST is unanchored at the bottom and no guidelines for the base overtuming momert of unanchored tanks are provided in the Generic Implementation Procedure, Revision 2 (GIP-2). Therefore, other guidelines and criteria were used in specific portions of the analyses of these tanks. However, the CSTis supported at the top and additionaljustification ofits seismic adequacy is discussed below. The details of the analyses, data and the references to support the criteria are contained in the calculations provided to the NRC Staff earlier. A summary of the methodology used in i
the various aspects of the analysis is provided below for these two tanks. The analyses of both tanks are based on sound technical guidelines and are systematically presented in the tank calculations. Criteria other than GIP-2 were used from well-known references and only where required and/or not provided in GIP-2. The use of these additional criteria does not conflict with those portions of the tank calculations where the GIP-2 methodology is used. The main references from which data and criteria were obtained for use in these calculations are also listed below.
Cendensate Storage Tank (CST):
The current analysis, which considers the CST as an unanchored tank (Calculation 52308.04-C-003) predicted that during the SSE only about 17" of the 28 ft diameter tank
. may get detached from the foundation with maximum vertical uplift of the tank base end less than 1". In this condition, it was calculated that the maximum compressive stress in the tank wall will be less than 70% of the critical buckling stress considering both elephant-foot and diamond-shaped buckling and the tank overturning moment capacity will meet the demand. The analysis predicted that the tank would not come in contact
with the concrete barrier during an SSE, even if the top angle support is ignored. For the various aspects of the analysis of this tank, the methodology we used is as follows:
. Sliding shear capacity: The calculation is based on GIP-2. Demand / Capacity < 0.5 Freeboard clearance: The calculatim uses the GIP-2 method. Demand / Capacity < 0.4 Convective (sloshing) mode response cf the tank: This is calculated using the GIP-2 method.
Impulsive mode response (base shear and overturning moment): The calculation uses the GIP-2 method.
Overturning Moment Capacity: Uplifting of the tank and the fluid hold-down forces were used to calculate the rnoment capacity of the tank. For unanchored tanks, the small displacement theory is excessively conservative, and an upper bound theory is considered appropriate, in accordance with Reference 2. The equations in reference 2 are based primarily on Flugge (ref. 3) and also on references 4 and 5. Reference 2 recommends limiting uplift height to 0.ll, where L is the uplift length of the base. The uplift length was estimated as 17" (less than 5% of the diameter), therefore,0.lL worked out as 1.7".
However, for additional conservatism, the uplift height was limited to 0.8" which is about 50% of the value recommended in reference 2.
Tank Stability: The criteria used for the tank buckling are primarily based on GIP-2.
Elephant-foot buckling follows the GIP-2 method, and the axial stress capacity is calculated from the equation provided in GIP-2, page 7-21, which is the same as EPRI
' NP-6041 (ref. 6). The diamond shaped buckling capacity is based on NASA SP-8007 (ref. 7), which is the same as the guidance provided in GlP-2.
Additional Justification for CST Seismic Adequacy (Note: This independently supplements the above-discussed recent analysis based on llNL and GIP-2 methodologies, which considered the CST as unanchored):
The CST is completely encased in 2' thick reinforced concrete (Reference 9 -
drawing 11715-FV-43A-6), which acts as a barrier for missile protection. There is a 2d space between the tank shell and concrete, which is filled with Rotofoam (a soft material) insulation. The insulation is provided to address any temperature effects of the fluid on concrete, but will also provide a limited amount of support to the tank and dampen seismic motions. The entire concrete shield and tank structure was seismically qualified as a unit in the original seismic design. However, the current seismic analysis of the tank treated the tank completely unanchored and did not take any credit for the reinforced concrete shield and the restraint provided by it. As shown in the drawings, there is an angle 2%"x 2%" x %" with one leg partially embedded to the concrete enclosure and welded to the dome, and the other leg welded all around to the top of the cylindrical portion of the tank. This support provides complete lateral restraint to the tank at the top. The pipe penetrations and
the manway are also reinforced with rings and pass through the concrete enclosure as shown in the attached drawings and sketches. Stone anal We'.ncr, the original Architect / Engineer for North Anna Power Station, took these seatures into account in the original seismic design of the tank. Therefore, even if the current analysis is disregarded, no gross overturning or sliding of the tank is possible because of the top support and the two feet thick concrete barrier. In addition, the top support j
Lwill alleviate catastrophic buckling of the tank, although our calculation, which uses i
- the methodology of BNL Report 52361, shows' that there is sufficient margin for huckling with the tank considered unanchored and unsupported at the top (see the response to Request b., which follows). Note that p 5-33 of BNL Report 52361 states that, " Tanks with lateral support near the top of the cylindrical tank wall are likely to have substuntially greater seismic capacity than a simikr tank without this top support."
The drawings and sketches for the CST are attached (Reference 9).
-l Refueling Water Storage Tank (RWSTl:
Tank responses,i.e., base shear and overturning moment including impulsive mode and convective (sloshing) mode: These calculations are primarily based on the GIP-2 methodology. Although in Attachment A of the calculation several equations, e.g. to calculate convective mode frequency, base shear and moment responses, are listed as being taken from Reference 6, they will provide the same result as the GIP-2 methodology.
Top plate stress, tank shell stress, vertical stiffener plates and chair-to-tank wall l
weld: The calculation uses GIP-2 equations.
Sloshing height and freeboard clearance: These calculations are based on GIP-2.
j Anchor bolt tensile capacity, oves turning capacity, and permissible uplift: The initial j
anchor bolt tensile capacity was based on GIP-2 and ACI 349. Subsequently, the capacity was revised by studying the failure mechanism for overturning as follows. The calculation of the overturning moment capacity, initially based on the GIP-2 criterion, showed that the moment capacity was lower than the demand because the capacity of the anchor bolt and the chair as a unit was not utilized. A further evaluation was performed based on a yield line analysis of the top plate of the anchorage chair that utilized ref. 8, included in Attachment E of the calculation. This analysis showed that the permissible load capacity of the anchorage was substantially higher than the initially calculated value, which did not account for the complete anchorage unit. Based on the revised anchorage I
capacity, the overturning moment capacity was reevaluated, as shown in Attachment D of the calculation. Note that the fluid hold-down forces were neglected, consistent with GIP-2. The permissible uplift was based on Ref. 2 (smaller than that pemlitted by GIP-
- 2).
l Tank Stability: The methodology used for evaluating tank buckling is primarily based on GIP-2. Elephant-foot buckling follows the GIP-2 method, and the compressive stress
k$
y 1
11
)
3
' t capacity is calculated from the equation provided in GIP-2. page 7-21, which is the same f
' as EPRI NP-6041 (ref.,6). The diamond shaped buckling capacity is based on NASA SP-
~ 8007._(ref. 7) and guidance from GIP-2.
I J
- Response to Request b.
The elephant foot buckling capacity (o ) and the diamond-shape buckling capacity (a a) pe p
for the CST were calculated using the GlP-2 (Reference 1) appmach. In this analysis, the -
' top restraint to the tank was conservatively ignored. _ In accordance with Step 16 of GIP-J2, p. 7-23, the allowable buckling stress was calculated as:
o = (0.9/1.25) * (miri(ope, o a)) = 0.72 Mn (ope, a a)) psi (1) c p
p
- This value is shown on the bottom of page 3 of 5 in Attachment B of the CST calculation'.
- 52308.04-C-003 submitted in response to the RAI on. April 1,' 1999. The value shown is Ca = 2.779 kip /in, which is the' maximum permissible axial compressive force per unit
' length and is obiained by multiplying the alloivable buckling stress c with the thickness e
L of the shell t, = 0.25"? The next steps in the calculation follow the approach of the DNL report (Ref. 2). An iterative procedure is used with MATHCAD. The compressive force
. per unit length (C%) at the outer compression side of the tank wall is calculated from 4
equation 5.21'of Ref. 2 and the moment capacity is based on equation 5.22 of Reference l
' 2. Both of these are functions of the circumferential angle to neutral axis and the uplift.
The angle is varied and the solution occurs at the circumferential angle of 2.855 radians
, and an uplift of 0.8" when the compressive force at the outer compression side of the tank
. wall equals the pennissible (allowable) compressive force, in accordance with step 4 on
. page 5-23 of Reference 2.
Since the pemiis'sible compressive force and the calculated compressive force are made equal, the above analysis method would imply that the factor of safety for buckling is
]
. unity.: However, based on the recommendation of Reference 2, an uplift displacement of
- up to 0.!L,'i.e.,1.77" is allowed (see bottom of page 4 of 5, Attachment B of Calculation 52308.04-C-003); The limiting uplift was only 0.8". Further, the evaluation showed that only 1.47' of the base'of the 28 diameter tank could potentially get detached from the foundatiori during overtuming caused by an SSE. When the applied overturning moment l equals the resisting moment, the. maximum compressive stress in the tank wall isLless 7
y than. 0.72 times the critical buckling stress, as evident by.use..of equation -(l) above.
s W"
. Additional; parametric ~ investigation was done and-it ' revealed that the maximum
- compressive stress is less than 0.6 times the critical buckling stress with a limiting uplift Tof about'1" (still much smaller than 0.lL), and the tank overturning capacity still meets 4
i a the demand.' Thus, the analysis shows that there is sufficient margin for buckling. - The f
' inverturning"of the tank, however, is not an issue due to the presence of the top support; cohsequently, the margin against buckling will be significantly higher than that presented 6
- m the current calculation.
n f
f 5
P
./
L ('
g
(
References:
- 1. Winston and Strawn et al., " Generic Implementation Procedure, Revision 2, (GIP-2)",
Seismic Qualification Utilities Group, February 14,1992.
- 2. Bandyopadhyay, K., et. al., " Seismic D'esign and Evaluation Guidelines for the Department of Energy. High --Level Wame Storage Tanks and Appurtenances",
Brookhaven National Laboratory, Report B'NL 52361, October 1995.
1
- 3. W. Flugge, Stresses in Shells, Springer-Verlag,1960.
^
- 4. G. C. Manos, "Earthqua'. Tank Wall Stability of Unanchored Tanks". Journal of Structural Engineering, Vol. I12, No. 8, ASCE, August 1986, pp 1863-1880.
)
1
- 5. M. A. Haroun and H. S. Badawi," Nonlinear Axisymmetric Uplift of Circular Plates",
Dynamics of Structures, ASCE, August 1987, pp77-89, t
- 6. Jack R. Benjamin Associates, et al., "A Methodology for Assessment of Nuclear i
- Power Plant Seismic Margin (Revision 1)", EPRI NP-6041-SL, August 1991.
- 7. " Buckling of Thin-Walled Circular Cylinders", National' Aeronautics and Space Administration, NASA SP-8007, August 1936.
- 8. R. P. Kennedy, Notes on yield line analysis", Structural Mechanics Consultmg, July 29,1986.
- 9. Drawings 11715-FV-43A-6,11715-FC-12.B-14 and sketches 720927-1 and 720921-1.
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i Relevant Pages from BNL Report 52361, dated October 1995
.