ML20059J897
| ML20059J897 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 01/26/1994 |
| From: | Kenyon T Office of Nuclear Reactor Regulation |
| To: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| References | |
| NUDOCS 9402010254 | |
| Download: ML20059J897 (19) | |
Text
.
'o UNITED STATES g
8 NUCLEAR REGULATORY COMMISSION o
h WASHINGTON, D. C. 20555
%.,*****/
January 26, 1994 Docket No.52-003 Mr. Nicholas J. Liparulo Nuclear Safety and Regulatory Activities Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, Pennsylvania 15230 s
Dear Hr. Liparulo:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON THE AP600 As a result of its review of the June 1992 application for design certifica-tion of the AP600, the staff has determined that it needs additional informa-the areas of structural engineering (Q220.24-Q220.50),information is needed in tion in order to complete its review. The additional and seismic design (Q230.24-Q230.49).
Enclosed are the staff's questions. Please respond to this request within 90 days of the date of receipt of this letter.
You have requested that portions of the information submitted in the June 1992 application for design certification be exempt from mandatory public disclo-sure. While the staff has not completed its review of your request in accordance with the requirements of 10 CFR 2.790, that portion of the submit-ted information is being withheld from public disclosure pending the staff's final determination. The staff concludes that this request for additional information does not contain those portions of the information for which exemption is sought. However, the staff will withhold this letter from public disclosure for 30 calendar days from the date of this letter to allow Westing-house the opportunity to verify the staff's conclusions.
If, after that time, you do not request that all or portions of the information in the enclosures be withheld from public disclosure in accordance with 10 CFR 2.790, this letter will be placed in the Nuclear Regulatory Commission's Public Document Room.
310021
- The numbers in parentheses designate the tracking numbers assigned to the questions.
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Mr. Nicholas J. Liparulo J noary ?6, 1994 This request for additional information affects nine or fewer respondents, and therefore, is not subject to review by the Office of Management and Budget under P.L.96-511.
If you have any questions regarding this matter, you can contact Kris Shembarger at (301) 504-1114.
Sincerely, (Original signed by K. M. Shembarger for)
Thomas J. Kenyon, Project Manager Standardization Project Directorate Associate Director for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enclosure:
See next page DISTRIBUTION:
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01/2[k9k 01/3(/94 Ol/D6/94 01/20/94 DATE 0FFICIAL RECORD COPY:
DOCUMENT NAME:
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Mr. Nicholas J. Liparulo Docket No.52-003 Westinghouse Electric Corporation AP600 cc: Mr. B. A. McIntyre Advanced Plant Safety & Licensing Westinghouse Electric Corporation Energy Systems Business Unit P.O. Box 355 Pittsburgh, Pennsylvania 15230 Mr. John C. Butler Advanced Plant Safety & Licensing Westinghouse Electric Corporation Energy Systems Business Unit Box 355 Pittsburgh, Pennsylvania 15230 Mr. M. D. Beaumont Nuclear and Advanced Technology Division Westinghouse Electric Corporation One Montrose Metro 11921 Rockville Pike Suite 350 Rockville, Maryland 20852 Mr. Sterling Franks U.S. Department of Energy NE-42 Washington, D.C.
20585 Hr. S. M. Modro EG&G Idaho Inc.
Post Office Box 1625 Idaho Falls, Idaho 83415 Mr. Steve Goldberg Budget Examiner 725 17th Street, N.W.
Room 8002 Washington, D.C.
20503 Mr. Frank A. Ross U.S. Department of Energy, NE-42 Office of LWR Safety and Technology 19901 Germantown Road Germantown, Maryland 20874 Mr. Victor G. Snell, Director Safety and Licensing AECL Technologies 9210 Corporate Boulevard Suite 410 Rockville, Maryland 20850 i
I I
REQUEST FOR ADDITIONAL INFORMATION
-i ON THE WESTINGHOUSE AP600 DESIGN i
STRUCTURAL ENGINEERING 220.24 Discuss the effects of wind-induced failure of non-safety-related structures on safety related structures, systems, and components (SSCs).
If the collapse of non-safety-related structures due to wind loading does not adversely impact the function of the safety-related SSCs, the use of 1.0 as the importance factor is suitable.
If not, an importance factor of 1.11 should be used in the design of such struc-tures. Therefore, the SSAR should provide a commitment that all SSCs not designed for wind loads should be analyzed using the 1.11 impor-tance factor or be checked that their mode of failure will not affect the ability of safety-related SSCs to perform their intended safety functions.
Provide that comitment or justify deviation from such a commitment (Section 3.3.1 of the SSAR).
220.25 At the transition region between the free-standing part and the encased portion of the steel containment, seals are providea at the top of the concrete at elevation 108 ft inside the vessel and at elevation 100 ft outside the vessel so that moisture is not trapped next to the steel vessel just below the top of concrete.
The seal on the inside accommodates radial growth of the vessel due to pressuriza-tion and heatup.
The staff is concerned about the mechanical proper-ties of this seal material and the stress conditions and buckling potential of the steel containment in this region.
No information is provided in the SSAR concerning (1) composition of the seal material, (2) the method used to obtain these material properties, (3) the uncertainties associated with these material properties, (4) the accessibility to perform periodic inspection, and (5) the behavior under the severe accident conditions. Address the issues associated with (1) the uncertainty of the mechanical properties of this seal material and the environmental qualification as well as age-related degradation management for the proposed 60-year design life for this seal material, and (2) the measures to be implemented to prevent collection of moisture in the transition region (Section 3.8.2 of the SSAR).
220.26 Provide information on measurements taken at critical locations during pre-operational structural integrity testing (SIT) of the steel containment. This information may be useful in validating the con-tainment analysis methods (Section 3.8.2 of the SSAR).
220.27 Provide the potential sources of a missile or sources of high pressure resulting from high-energy line break between the steel containment and the operating floor and refueling cavity walls, between the secondary shield walls and the steel containment, and between the steel containment and the shield building (Section 3.8.2 of the SSAR).
Enclosure
i 4 '
220.28 Since the air baffle is supported at the top of the shield building and is attached to the steel containment through 3" p pipe supports, discuss what considerations were given to the design of.the air baffle, and the effect of the baffle on the steel containment and shield building for all the loading conditions, specifically the seismic loads and the severe accident loads (thermal and pressure).
Describe in more detail (possibly with figures) the flexible seal at the top of the air baffle and the connection to the shield building roof (Section 3.8.2 of the SSAR).
220.29 Overall seismic loads result in axial compression and tangential shear-stresses which are greatest at the base of the cylindrical portion of the containment. Westinghouse evaluated the shell for dead load, live load, and seismic load at the critical section close to the _ bottom tangent line. Westinghouse reported that the calculated stress was 2721 psi and the corresponding allowable stress for the Level C Service Limit was 4438 psi based on NE-3133.6. Axial and tangential shear stresses were evaluated in accordance with the ASME Code Case N-284.
The maximum value of the interaction ratio was reported as 0.5 and the allowable interaction ratio was 1.0.
However, Westinghouse combined seismic loads by the (1.0, 0.4, 0.4) method and added to the dead load and live load.
Provide the basis for the use of only this method in combining seismic loads and not the SRSS method as described in Section 3.7.2.6 (Section 3.8.2 of the SSAR).
220.30 Westinghouse estimates the maximum pressure at ambient temperature corresponding to the following stress and buckling criteria:
(1) deterministic severe accident pressure capacity corresponding to ASME level C Service Limit on stress intensity, Code Case N-284 for buckling of the equipment hatch covers, and two-thirds of critical buckling for the top head, and (2) best estimate capacity correspond-ing to gross membrane yield at the ASME-specified minimum yield stress (SA 537, Class 2, yield stress - 60 ksi, ultimate stress - 80 ksi),
and critical buckling for the equipment hatch covers and top head.
However, neither the' Code Case N-284 for buckling of the equipment hatch covers (see Q220.32) nor the two-thirds of critical buckling for the top head is acceptable. The factor of safety due to the internal pressure (Appendix A) is 1.67 for the Level C Service Limit as speci-fled in the Code Case N-284. Note (1) in Table 3.8.2-2 should be revised to reflect that the factor of safety is 1.67, or acceptable justification should be provided for not doing so.
In addition, Westinghouse analyzed the steel containment vessel for the theoretical buckling capacity using the BOSOR-5 computer code, which uses both large displacement and nonlinear material properties.
The yielding started at a pressure of 144 psig for the cylinder, at 146 psig for the top of crown, and at 152 psig for the knuckle region, using elastic-plastic material properties, a yield stress of 60 ksi, and the von Mises yield criterion.
Provide the bases for the use of von Mises criterion instead of ASME stress intensity criterion to establish yield.
\\
l l -
Westinghouse determined that the theoretical plastic buckling pressure is 174 psig. At the pressure of 174 psig, Westinghouse calculated the maximum effective pre-buckling strain of 0.23 percent in the knuckle i
region where buckling eventually occurred, and 2.5 percent at the crown. However, it is not clear how these strains were derived.
For the SA 537 Class 2 material, it is reported that the stress-strain curve has the strain plateau from 0.2 percent to 0.6 percent without pressure increase and strain hardening after 0.6 percent (see Sec-tion 3.8.2.4.2.6 of the SSAR). At the knuckle region, Westinghouse states that it started to yield at the pressure of 152 psig, which will go to 0.6 percent strain with no further pressure increase. At the top of the head, the expected stress at 174 psig is 72 ksi and the corresponding strain is about 8 percent.
Explain how a value of only 2.5 percent strain was obtained and provide justification for the ultimate capacity of the containment (Section 3.8.2 of the SSAR).
220.31 In SECY-93-087, " Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light Water Reactor Designs," the staff proposed that the containment be evaluated for the credible severe accidents against the stress limits of the ASME Level C Service Limit.
Westinghouse states that for tensile stresses at the cylindrical portion of the containment, this results in a pressure capacity equal to 125 psig by ASME stress intensity criterion. However, the staff estimated it as 114 psig using ASME stress intensity criterion based on the theoretical stress calculations.
Explain this discrepancy (Section 3.8.2 of the SSAR).
i 220.32 Westinghouse estimates critical buckling pressures for equipment hatches as 196 psig for a 22-foot-diameter hatch and 161 psig for a 16-foot-diameter hatch. The corresponding ASME Level C Service Limits are 117 psig and 96 psig using the Code Case N-284, respectively.
From Figure 3.8.2-2, the equipment hatch covers appear convex to the center line of the containment. Therefore, the use of the Code Case N-284 (i.e., the factor of safety of 1.67 for the Level C Service Limit) is not acceptable because the internal pressure of the contain-ment acts as the external pressure to the spherical cap covers and subjects the cap covers to compression.
In the case of external pressure, ASME NE-3222 (i.e., the factor of safety of 2.5 for the Level C Service Limit) should be used for the compressive stresses.
Note (1) in Table 3.8.2-2 should be revised to reflect the factor of safety of 2.5, or acceptable justification should be provided for not doing so (Section 3.8.2 of the SSAR).
220.33 NUREG/CR-5334 reported that, during severe accident conditions, no l
1eakage was detected from any of the three current electrical penetra-tion assemblies (EPAs), under the following conditions (1) D. G.
O'Brien EPA, 361'F, 155 psia for 10 days, (2) Westinghouse EPA, 400*F,
- Timoshenko, S. and Woinowsky-Krieger, S., Theory of Plates and Shells, pp 484-485, second edition, 1959, McGraw-Hill.
d
. 75 psia for 10 days, and (3) Conax EPA, 700*F, 135 psia for 10 days.
However, the SSAR does not address what EPAs will be used for the AP600.
Provide a commitment in the SSAR that EPAs penetrating con-tainment be at least as strong as the steel containment vessel (Sec-tion 3.8.2 of the SSAR).
220.34 Nonmetallic items, such as gaskets, are qualified to function at the design temperature. The SSAR should provide the functionality of such items under the severe accident conditions (Section 3.8.2 of the SSAR).
220.35 Provide a corrosion allowance to be used for the proposed 60-year plant design life and its technical basis. Also, indicate whether post-weld heating during construction is provided for the steel containment plates (Section 3.8.2 of the SSAR).
220.36 Submit the stress analysis results for the most highly stressed portions of the containment shell in both meridian and circumferential directions (Section 3.8.2 of the SSAR).
220.37 Submit the pre-buckling stresses for the most highly stressed regions in both meridian and circumferential directions, and verify that stresses at buckling are within the elastic range (Section 3.8.2 of the SSAR).
220.38 Discuss wnether all strains in the axisymmetric analysis model are i
comparable to the Sandia strain criteria" (Section 3.8.2 of the SSAR).
220.39 Discuss whether the strains at all discontinuities (i.e., around penetrations and penetr,ation reinforcements) are comparable to the Sandia strain criteria (Section 3.8.2 of the SSAR).
220.40 Discuss the effects of concrete cracking for the seismic analysis of-all AP600 Category I structures (Section 3.8.3 of the SSAR).
220.41 Discuss the design of the embedded portion of the exterior walls of the nuclear island of seismic Category I_ structure and the methods for l
the consideration of static soil pressure and the soil pressure induced by the earthquake. Westinghouse should follow the guidelines documented in the staff position for the embedded wall and retaining wall design.
Evaluate the potential local soil failure around the embedded walls during the design seismic event (Section 3.8.4 of the SSAR).
" Miller, J. D. and Clauss, D. B., " Evaluation of the Performance of the Sequoyah Unit 1 Containment Under Conditions of Severe Accident loading,"
NUREG/CP-0095, Paper No. SAND 88-1631C, pp 571-588, 1988.
220.42 Westinghouse states that for all safety-related structures, the design rainfall is 493 mm/km2/hr (19.4 in/mir/hr). The roof of the seismic Category I Structures should be designed to have parapets with scup-pers to supplement roof drains or be designed without parapets so that excessive ponding of water cannot occur.
Provide detailed design criteria against severe weather phenomena, such as heavy rainfall and snow loadings (Section 3.8.4 of the SSAR).
220.43 The applicant for an combined construction / operating license (COL) will need to ensure that the settlement of adjacent buildings will be such that the integrity of underground piping or tunnel will not be jeopardized. The SSAR should contain a statement that the COL appli-cant should perform stability evaluations of all safety-related facilities, including foundation rebound, settlement, differential settlement, and bearing capacity.
Provide that statement (Sec-tion 3.8.4 of the SSAR).
220.44 State which methodology (SRSS or (1.0, 0.4, 0.4) method] is used for the seismic loads calculation.
For the computation of global seismic loads, indicate whether the inertial properties include all tributary mass expected to be present in operating conditions at the time of earthquakes. This mass should include the dead load, stationary equipment, piping, and appropriate part of the live load (Sec-tion 3.8.4 of the SSAR).
220.45 Provide a commitment to design all subcompartments for global pres-sure/ temperature effects, and provide the actual pressure / temperature values to be used (Section 3.8.4 of the SSAR).
220.46 Specify whether epoxy-coated reinforcing steel is used for areas where a corrosive environment is encountered (Section 3.8.4 of the SSAR).
220.47 Specify the analysis methods and the design criteria for seismic-Category II structures (Section 3.8.4 of the SSAR).
220.48 Identify the particular areas that raise the concern about the capa-bility of connection, reinforcement pattern or welded joint (Section 3.8.4 of the SSAR).
220.49 The seismic Category II structures, such as the turbine building, the annex buildings I and II, and the solid radwaste building are suffi-ciently close to the nuclear island such that their collapse could affect the safety function of Category I structures. The structural integrity is the requirement for seismic Category II structures.
Therefore, provide the reason why the seismic Category II structures are excluded for the foundation analyses (Section 3.8.5 of the SSAR).
220.50 The factor of safety against sliding and overturning the nuclear island due to tornado and wind should be provided.
In Table 3.8.5-1, provide the rationale for the buoyancy force criterion for the sub-4 merged structure (Section 3.8.5 of the SSAR).
SEISMIC DESIGN 230.24 Section 3.7 of the SSAR states that a three-level seismic classification system is used for the AP600; seismic Category I, seismic Category II, and non-Category I.
However, Section 3.2.1 (Seismic Classification) of the SSAR states that the methodology classifies. structures, systems and components into three categories:
seismic Category I (C-1), seismic Category II (C-II) and non-seismic (NS). Clarify the difference between non-Category I and non-seismic.
230.2S Section 3.7 of the SSAR states that non-Category I structures are designed or physically arranged (or both) so that the safe shutdown earthquake (SSE) could not cause unacceptable structural interaction with or failure of seismic Category I structures, systems and components.
However, Section 3.7.2 of the SSAR states that seismic Category II structures are designed and/or physically arranged so that the SSE could not cause unacceptable structural interaction with or failure of seismic Category I structures, systems and components.
These two statements imply that classifications for non-Category I and seismic-Category II are the same. Clarify these statements.
230.26 Page 3.7-1 of Section 3.7 of the SSAR states that the AP600 standard plant used a three-level seismic classification, i.e., seismic Category I, seismic Category II and non-Category I.
Section 3.7 specifies the general design requirements for the seismic Category I items.
It also specifies the general design requirements for the non-Category I items. However, the general design requirements were not provided for the seismic Category II items.
Provide this information.
230.27 On Page 3.7-1 of Section 3.7.1.2, the last paragraph states that SRP Section 3.7.1 contatiis the provision of frequency intervals used in the computation of the response spectra. Was this SRP provision satisfied in the computation of the response spectra?
230.28 a.
In Section 3.7.1.2 of the SSAR, the cross-correlation coefficients between the three components of the ground motion time history should be specified to demonstrate that these three components are statistically independent.
Provide that information.
b.
Provide the procedures for the development of the vertical target PSD in Section 3.7.1.2 of the SSAR.
c.
Explain the meaning of "with 20% averaging," as shown in Figures 3.7-10 through 3.7-12 of the SSAR.
230.29 On Page 3.7-2 and in Table 3.7-1 of the SSAR, the damping ratios assigned for HVAC ductwork, cable trays and fuel assemblies are 7%,
20% and 20%, respectively.
Provide the bases for these parameters to justify the adequacy of using high damping ratios for the analyses of the welded ductworks, cable trays and fuel assemblies.
230.30 Figures 3.7.1-14 and 3.7.1-15 of the SSAR provide the damping values for rock material and soil material, respectively, a.
Clarify what damping values are to be used for hard-rock material and soft-rock material.
b.
Provide the basis and source of these two figures.
-l 230.31 Section 3.7.1.4 of the SSAR describes the shear wave velocity profile for the supporting media from ground surface to a depth of 240 ft for both soft-rock site condition and soft-to-medium stiff soil site condition, and states that the base rock is at a depth of 120 ft.
a.
What is the shear wave velocity profile for the base rock?
b.
Because the base rock is at the depth of 120 ft, how significant will it be to specify the shear wave velocity profile from the depth of 120 ft to 240 ft for both soft-rock and soft-to-medium stiff soil sites? Provide such a profile or provide justification for not doing so.
c.
Because the location of the base rock is not shown in Figure 2A-7
)
of SSAR, provide a complete plot for the shear wave velocity profile for hard-rock, soft-rock and soft-to-medium stiff soil
{
sites.
230.32 In Section 3.7.1 of the SSAR, the location of the input ground motion to be specified is not shown for the site conditions selected.
Provide this information.
230.33 From the staff's review of Section 3.7.1.4 and Appendix 2A of the
'l SSAR, it appears that only three design soil profiles are required for the design of AP600 seismic Category I structures, and that some potential governing site conditions, such as a shallow soil site and a deep soil site, were not considered.
Provide justification to demon-strate that the design of the seismic Category I structures and subsystems based on these three site conditions can envelop the design of the structures and subsystems founded on other potential sites in the United States.
230.34 The term " time history analysis" appears to be inconsistently used i
throughout Section 3.7.2 of the SSAR.
In some cases, it is mixed with the term " complex frequency response analysis." From the staff's review of Section 3.7.2.1.2 of the SSAR, it is the staff's understand-ing that the (modal) time history analysis method was used for the fixed base structural model (hard-rock site condition) to generate floor response spectra, and the complex frequency response analysis method was used for the soil-structure interaction analysis when the structures are founded on soft-rock site and soft-to-medium stiff soil site.
Is this correct? Clarify any inconsistency.
l
230.35 The following request for additional information pertains to Section 3.7.2.1.1 of the SSAR.
i a.
Provide the detailed comparison of the results obtained from the 2D SSI analyses and the 3D response spectrum analyses for the hard-rock site condition.
~
b.
As described in Section 3.7.2.1.1, the structural member forces and moments are obtained from the response spectrum. analysis of
'i the finite element model for the hard-rod site, and from the SSI analysis of the stick model for the soil sites.
Provide a compar-ison of responses frora the response spectrum analyses of a stick model and a finite element model at rock site.
c.
From the staff's review of Section 3.7.2.1.1 and Table 2A.17, the staff determined that the hard-rock site condition (RI) is not the governing case for the steel containment shell.
Describe how the steel containment shell was analyzed for the rock site condition.
d.
Provide the rationale for excluding the SB roof in the finite element model, as shown in Figure 3.7.2-1.
e.
From the staff's review of Tables 3.7.2-1 through 3.7.2-4 of the SSAR, the staff determined that the AP600 nuclear island struc-tures (except the steel containment shell) are very rigid.
Some predominant frequencies are much higher than 33 Hz.
Provide justification for the statement "since the shear wave velocity for the hard rock site is.in excess of 8000 ft per second, the soil-structure interaction effect is negligible." This statement has also been made in Sections 3.7.2.1.2 and 3.7.2.4.
230.36 The following request for additional information pertains to Section 3.7.2.1.2 of the SSAR:
a.
Provide the validation package of computer code SASSI for review.
b.
Explain the difference between the two phrases " applied sim-ultaneously (time history in the seismic analysis)" and " applied separately."
230.37 Section 3.7.2.2 of the WR describes the importance of the mass participation from the high frequency structural modes of the stick model in the horizontal and, particularly, the vertical directions due to the rigidity of the containment internal structures.
It is not clear how the contributions of the predominant high frequency modes to, the structural responses were taken into account in the analyses.
Particularly, provide the following information:
a.
The cutoff frequencies used in the horizontal and vertical time-history analyses of the fixed base model (the case of structures founded on rock site).
_g_
b.
The cutoff frequencies used in the horizontal and vertical SSI analyses using the complex frequency response analysis method, and provide the basis for the cutoff frequencies selected.
c.
Details of the separate seismic analysis using the coupled con-tainment internal structures and reactor coolant loop (RCL) lumped-mass model (Page 3.7-5, first paragraph) and the difference in the response results between this separate seismic analysis and the original seismic analysis. Was this " separate analysis" done using the fixed base model for the rock site condition?
d.
Details for considering the high frequency effect to the vertical responses (forces and moments) of the containment internal struc-tures (Page 3.7-5, first paragraph). Was this consideration applied only to the vertical seismic analysis of the fixed base structural model for the rock site condition't 230.38 Provide a description in the SSAR to show how the other seismic Category I structures such as containment air baffle (CAB), passive -
containment cooling system water storage tank (PCCSWST), and in-containment refueling water storage tank (IRWST) were included in the nuclear island seismic models (lumped-mass stick model and finite element model) (Section 3.7.2.3 of the SSAR).
230.39 The following request for additional information pertains to Section 3.7.2.3.1 of the SSAR:
a.
Explain how the live loads were considered in the modeling of the coupled shield and auxiliary buildings and the containment inter-nal structures.
b.
In the second paragraph of Page 3.7-6, the SSAR states that two sticks were used to represent each structure (shield building, auxiliary building or containment internal structures). The first stick represents the axial areas and the second stick represents the beam element properties other than the axial areas.
It seems i
that this modeling technique is trying to decouple (a) the axial and bending responses, and (b) the horizontal and vertical responses.
Explain and justify this modelling technique.
c.
If the containment internal structures are represented by two separate sticks, explain how to couple the RCL model with the internal structural model.
230.40 Section 3.7.2.3.2 of the SSAR states that the three-dimensional, lumped-mass stick model of the steel containment vessel is developed based on the axisymmetric shell model. This implies that the steel containment shell model is axisymmetric. However, Section 6.3.2.2.3 of the SSAR states that the in-containment refueling water storage tank (IRWST) is constructed as an integral part of the containment structure.
In addition, when the polar crane is included in the
i dynamic model, the trolley should be assumed to be parked at the end of the crane girder.
Explain how the steel containment shell can be modeled from an axisymmetric shell model.
230.41 a.
Section 3.7.2.3.3 of the SSAR states that for soil-structure interaction (SSI) analyses, the nuclear island basemat and the periphery walls of the embedded portion of the nuclear island are represented by a three-dimensional finite element model. When the basemat was modeled, has the flexibility of the basemat been considered in the SSI analyses?
b.
Evaluate the possibility of the out-of-phase interaction between the shield building, steel containment vessel and containment air baffle (Section 3.7.2.3.3 of the SSAR).
230.42 The following request for additional information pertains to Section 3.7.2.4 of the SSAR:
a.
The first paragraph of Section 3.7.2.4 states that the nuclear island SSI responses generated for the analysis and design of seismic subsystems include nodal displacements, nodal accelera-tions and floor response spectra (FRS).
Explain how the structur-al member forces (forces and moments) used for the structural design were generated for a soil site condition.
b.
The last paragraph of Section 3.7.2.4 (Page 3.7-7) states that the selected soil conditions envelop the potential variation of soil i
properties and, therefore, the guidelines of SRP Section 3.7.2 for the variation of soil properties were not considered.
Justify this statement, especially, when structures are founded on soft soil site for which the variation (uncertainty) in soil properties should be carefully considered, c.
Explain the differences between the two phrases'"the time-history SSI analysis using the program SASSI" and "the complex frequency response analysis using the program SASSI."
3 d.
When the computer code. SHAKE was applied,~which soil degradation curve was used?
230,43 a.
The second paragraph of Section 3.7.2.5 states that seismic floor response spectra are computed using the nodal time-history responses determined from the nuclear island seismic time-history analyses with the various design soil profiles. As stated in Section 3.7.2.1.2, the complex frequency response analysis using the computer program SASSI was applied for the soil site condi-tions to generated the FRS.. Clarify this statement.
b.
The second paragraph of Section 3.7.2.5 states that Fig-ures 3.7.2-24 through 3.7.2-26 present the safe shutdown earth-quake FRS for the hard rock site condition at selected locations i
_ 11 of the coupled model of the shield and auxiliary buildings, the steel containment vessel and the containment internal structures.
Provide the FRS for the other site conditions.
230.44 When the complex frequency response analysis method was used, were the three components of the earthquake motion applied simultaneously or separately (Section 3.7.2.6 of the SSAR)?
~230.45 The second paragraph of Section 3.7.2.7 of the SSAR states that in the time-history analyses, combination of modal responses is not neces-sary.
It is not clear how the modal time-history analyses (using the computer program BSAP) for a fixed base structural model (structures founded on hard rock site) were performed. Clarify this statement.
230.46 Section 3.7.2.11 of the SSAR states that the seismic analysis models of the nuclear island incorporate the mass and stiffness eccentrici-ties of the seismic Category I structures and the torsional degrees of freedo:n and, hence, additional accidental torsion is not added to the actual calculated torsional responses. According to SRP Section 3.7.2, to exclude the accidental torsion to the overall seismic responses is not acceptable to the staff.
Provide justification for this deviation to the SRP.
230.47 Section 3.8.2.1.2 of the SSAR states that the vertical and lateral loads on the containment vessel and internal structures are trans-ferred to the basemat below the vessel by friction and bearing. This statement implies that there are no shear studs or anchors between the internal structures, steel containment vessel and reinforced concrete basemat.
Provide an analysis to demonstrate the dynamic stability of the containment vessel during an SSE event or a seismic margins earthquake.
230.48 Provide a detailed description regarding the " design by rule" analysis method in the SSAR and discuss what activities are underway for adoption of this method by a consensus code or standard (Sec-tion 3.7.3.1 of the SSAR).
230.49 The following request for additional information pertains to Sec-tion 3.7.3.3 of the SSAR:
a.
As described in Section 3.7.3, the structural frames and miscella-neous steel platforms are also considered subsystems.
Provide detailed modeling procedures and the analysis methods used for these substructures.
b.
Were the modelling procedures described in Section 3.7.3.3 also used for modelling the cable tray, HVAC and conduit systems?
Clarify this section.
c.
If some safety related piping systems and/or components are supported by those structural frames and miscellaneous platforms described in (a) above, discuss in detail:
(1) how the structural
frames and platforms were modeled together with the piping systems and components, and (2) how the potential amplification of motion through these frames and platforms were considered or are to be considered.
d.
Discuss how the polar crane system was modeled, analyzed and designed.
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Appendix A I
ECGB Position on Shell Buckling due to Internal Pressure INTRODUCTION Generally, when people speak of shell buckling, they refer to the buckling of i
the shell under external pressure. Therefore, the first reaction to most l
people to the suggestion that internal pressure, in a shell container, can cause buckling is skeptical. This is quite understandable, because from our experience with the design of spheres and cylinders closed by hemispheres, for instance, as containment vessels in nuclear power plants, the membrane stresses in these shells are tensile when subjected to internal pressure.
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l Containment vessels with the above-mentioned configurations are assessed for buckling due to potential external pressure.
In addition considerations are given to compressive and shear membrane stress fields, which can occur in containment shells during earthquakes or an internal asymmetric pressure due j
to LOCA or a variant distribution of pressure around the circumference. This results in axial compression in some portions of the containment shell and shear across the shell section. Compressive hoop stress can also occur at the point of support of a containment vessel under internal pressure where the l
movement of the shell is restrained.
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Because of their geometrical confiourations, torispherical and ellipsoidal shells under internal pressure have a stress field in which membrane tension 1
in the meridian direction and compession in the hoop direction exist with the potential for buckling if not properiy designed. This can be shown theoretically and has been demonstrated experimentally. Most of the steel containments for pressurized water reactor (PWR) plants in the U.S. are of spherical or cylindrical with hemispherical dome configurations.
For boiling water reactor (BWR) plants, steel containment configurations vary from inverted bulb surrounded by a torus, cylinder topped by a conical frustum to a cylinder with an ellipsoidal shallow dome.
It appears that all the drywell heads in BWR plants are of torispherical configuration. The steel containments and their appended steel components are designed in accordance with the requirements of the ASME codes acceptable to NRC at the time of the licensing application.
To determine the viability of the containment during a' reactor severe accident including a core-melt, it becomes necessary to know the ultimate capacity of the containment more precisely and with a margin of safety.
This can be observed from item III.D, Containment Performance, contained in the enclosure to SECY-90-016 which states:
The containment should maintain its role as a reliable leaktight barrier by assuring that containment stresses do not exceed ASME Level C Service Limits for a minimum period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the onset of core damage and that following this 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period the containment should continue to provide a barrier against the uncontrolled release of l
fission products.
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This requirement appears to be applicable only to steel containment and its appended components which are under internal pressure and it does not mention how buckling is to be considered. This is because buckling is generally perceived to be a problem mainly for thin shells under external pressure as evidenced by the requirements stipulated in the ASME Section III Subsection NE Code.
In view of these facts it is essential that a rational criteria be established for evaluating buckling of shells under internal pressure.
In the following the criteria for buckling as contained in ASME Section III Subsection NE and in Code Case N-284 are first examined to discern the relation between the two and the basic philosophy behind them. On the basis of this understanding supported by the extensive theoretical and experimental studies available in the literature, it is believed that a determination can be made whether the stipulations in the NE sections or the Code Case N-284 on buckling can be applied to the buckling of the shell under internal pressure or separate new criteria need to be established.
REVIEW 0F ASME BUCKLING CRITERIA Subsection NE:
The design of the steel containment against buckling is based on requirements contained in NE-3133 and in NE-3222.
NE-3133 gives formulae to determine the allowable external pressure for different shell configurations.
The external pressure thus determined is implied to have already included factors of safety and capacity reduction factors. NE-3222.1 specifies the allowable values for the basic compressive stress which may arise from mechanical, thermal and pressure loads. The basic maximum buckling stress values to be used for the evaluation of stability are to be either (a) one-third of the value of the critical buckling stress determined by one of the following methods: (1) rigorous analysis considering all effects which can influence buckling, (2) classical analysis reduced by margins (knockdown factors), and (3) model testing, or (b) the value obtained from NE-3133. NE-3222.2 stipulates stability stress limits in percentages of the value given in NE-3222.1 as follows: (1) for design conditions and Level A and B Service Limits use 100%, (ii) for Level C Service Limits use 120% and (iii) for Level D Service limits use 150%, which can be translated into factors of safety of 3, 2.5 and 2 respectively for NE-3222.1 (a).
NE-3324.4 and NE-3322.6 provide the formulae for determining the thicknesses of ellipsoidal and torispherical heads respectively for internal pressure with limitations on radii to avoid compressive stresses. The thicknesses are determined on the basis of Levels A and B Service Limits.
Code Case N-284:
The purpose of this Case is to provide stability criteria for determining the structural adequacy against buckling of containment shells with more complex shell geometries and loading conditions than those covered by NE-3133.
Even though the Case lists a number of complex conditions, a careful reading of the case will lead one to conclude the 2
case applied basically to local buckling of stiffened and unstiffened shells under external or internal pressure, stringer buckling and general instability of stiffened shell under external pressure. The basic compressive allowable stress values referred to by NE-3222.1 will correspond to a factor of two in this Case. The stability stress limits referred to by NE-3222.2, in this Case will correspond to the following factors of safety:
2, 1,67 and 1.34 respectively for the three conditions of Service Limits as indicated under NE-3222.1 (a). These factors of safety are the minimum values required for local buckling. The respective factors of safety for stringer buckling and general stability failures are required to be 20% higher than those for critical local buckling, that is, the factors of safety to be applied are 2.4, 2.0 and 1.6 for the three conditions of Service Limits. It is to be noted that in addition to the' factors of safety, capacity reduction factors which account for the effects of imperfections and nonlinearity in geometry and boundary conditions and plasticity reduction factors which account for nonlinearity in material properties are to be applied in accordance to the guidance given in the code case.
Further it should be mentioned that Code Case N-284 has been endorsed in Regulatory Guide 1.84 Revision 27 with a condition relating to the consideration of the effect of the presence of a large opening on the shell.
From the above it can be stated that Code Case N-284 is a supplement to NE-3133 and NE-3222 and takes into consideration of the local buckling of the shell whether stiffened or unstiffened, of the stringer buckling and general stability of the stiffened shell as a whole. Such consideration is lacking in either NE-3133 and NE-3222. As observed from above, the factors of safety for local buckling whether stiffened or unstiffened, for stringer buckling and for general stability failure of stiffened shell.are smaller than those for the general buckling of unstiffened shells. This is due to the fact that with stiffeners the shell is less sensitive to imperfections and the stiffened shell has a higher resistance against buckling.
Furthermore, local buckling of the shell whether stiffened or.unstiffened has no effect on the stability of the shell as a whole. Therefore, it would be unnecessarily conservative to use the factors of safety as specified in NE-3222 for general stability of the shell for local buckling.
Code Case N-284 states that the basic facter of safety of two is applied to buckling stress values that are determined by classic (linear) analysis which has been reduced by capacity reduction factors determined from lower bound values of test data.
It is to be noted when Code Case.N-284 is applied to shells under internal pressure, the influence of the internal pressure may reduce the initial imperfections and therefore higher values of capacity reduction factors may be used.
CRITERIA FOR SHELL BUCKLING DUE TO INTERNAL PRESSURE From the above review and observation we can conclude that shell buckling due to internal pressure, as in the case of ellipsoidal and torispherical shells, 3
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should be evaluated on the basis of ASME Code Case N-284 as local buck',ing.
This is because the buckling of such shells under internal pressure i!. of the stable kind. After the first one or two buckles have formed it is possible to keep on increasing the internal pressure with additional buckles appearing periodically but there is no effect on the stability of the overall shell.
However, it should be noted, that the formation of these circumferential buckling waves on the shell can fracture the joints with any components appended to this portion of the shell and damage the components such as bellows, closure of openings, and other attachments. There is also the possibility of the fracture of the shell wall itself in the process of forming the buckles, if the shell steel material is brittle. With the continuous increase in the internal pressure and after the cessation of the formation of buckles without any fracture, the shell will most likely fail by axisymmetric yielding.
CONCLUSION On basis of a careful review and evaluation of the NE sections and Code Case N-284 on buckling it is recommended that the buckling of ellipsoidal and torispherical shell due to internal pressure should be considered as local bucklina and evaluated ch the basis of the criteria as contained in Code Case N-i"'.
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