ML19317H029
| ML19317H029 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 07/31/1968 |
| From: | Hall W, Newmark N, Webb Patricia Walker NATHAN M. NEWMARK CONSULTING ENGINEERING SERVICES |
| To: | |
| Shared Package | |
| ML19317H023 | List: |
| References | |
| NUDOCS 8004030710 | |
| Download: ML19317H029 (12) | |
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.i R220RT TO A2C 22GULATORY ST!.77 AD2G. 0Y 07 TdE STRUCTURAL CRIT 2RIA 703 Td7.-
RANCHO 0200 NUCL2AR GIN 2 RATING STATION UNIT NO. 1 Sacramento Municipal Utility District (AIC Docket No. 50-312) s bY t
N. M. Newmark W. J. Hall W. H. Walker A. J. Hendron f
i July 1968 9
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m ADEQunCY 07 THE STRUCTURAL CRITERIA fu THE RANCHO SECO NUCII.AR GENERATING STATION UNIT NO.1 by 1
N..M. Newmark, W. J. Hall, W. H. Walker and A. J. Hendron INTRODUCTION This-report is concerned with the adequacy of the containment structures and' components for the Rancho Seco Nuclear Generating Station Unit No. 1 for which application for a construction permit has been made to the U. S. Atomic Energy Commission by the Sacramento Municipal Utility District. The facility
.is located in Sacramento County, California, approximately 26 miles NNE of Stockton and 25 miles SE of Sacramento, California.
Specifically this report is concerned with the design criteria that determine the ability of the. containment system and Class I equipment and piping, as well as Class II structures and equipment, to withstand an I
Operating Losis Earthquake of 0.13g maximum transient horizontal ground
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acceleration simultaneously with the other applicable loads forming the
, basis of the design. The facility is also to be designed to withstand a Design Basis Earthquake of 0.2Sg maximum horizontal grcund acceleration to the extent of insuring safe shutdown and containment.
This report is based on information and criteria set forth in the Preliminary Safety Analysis Report (?SAR) and amendments thereto as listed 1
at the end of this report. Also, we have participated in discussions with the applicant and the AEC Regulatory Staff concerning the design of this unit.
DESCRI? TION 07 FACILITY The kancho Seco Nuclear Cencrating Station Unic No. 1 is described in
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the. ?SAR as consisting of a pressurized-water type reactor employing two closed 1
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g Each reactor cooling loops connected in parallal to the reactor vessel.
The nuclear steam supply _ system coolant loop is connected.co a steam generator.
. ill be furnished by the Babcock and Wilcox Company, and the turbine generator w
11s to be supplied by the Westinghouse Electric Corporation. The plant is to be designed for a power level of 2452 MWt (827 MWe).
The containment structure is a fully continuous reinforced concrete structure in the shape of a cylinder with a shallov domed roof and an essentially flat foundation slab. The cylinder has an internal diameter of 115 ft., and an inside height of 205 ft.
The distance from the top of the i
foundation slab to the springline of the domed roof is approximately 166 ft.
The vertical wall thickness is noted to be 3 ft.-9 in., and the dome thickness, 3 ft.-3 in.
The foundation slab thickness is approximately 10 ft.
The cylindrical portion of the cor.cai. ment structure is prestressed by a post-ten.ioning system consisting of horizontal and vertical tendons. The hoop tendons ara placed in a 3-120 system using 6 buttresses as anchorages; the
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dome has a 3-way post-tensioning system. The foundation sla'o is reinforced with conventional reinforcing steel. The inside of the containment struc-ture is lined with a steel plate noted to be 1/4 in. chick in the drawing contained in the PSAR.
The design of the contdinment structure' for this facility is similar to that employed for the Florida Power and Light Company's Turkey Point Plant, and Consumers Power Company's Palisades Plant.
Three general types of steel prestressing tendon systems have been under
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considaration by the applicant, namely those designated as Freyssinet, 33RV, and Stress Steel (S2EE). The applicant has stated thae the B32V tende, h
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Each reactor cooling loops connected in parallel to the reactor vessel.
The nuclear steam' supply system coolant loop is connected to a steam generator.
will be furnished by the Babcock and Wilcox Company, and the turbine generator
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is to be supplied by the Westinghouse Electric Corporation. The plant is to be designed for a power level of 2452 MWt (827 MWe).
The containment structure is a. fully continuous reinforced concrete structure in the shape of a cylinder with a shallow domed roof and an essencially flat foundation slab. The cylinder has an internal disaccer of 116 ft., and an inside height of 205 ft.
Tha distance from the top of the foundation slab to the springline of the domed roof is approximately 166 ft.
The vertical wall thickness is noted to be 3 ft.-9 in., and the dome thickness, 3 ft.-3 in.
The foundation slab thickness is approximately 10 ft.
The cylindrical portion of the contai.mont structure is prestressed by a post-ten.ioning system consisting of horizontal and vertical tendons. The hoop tendons are placed in a 3-120 system using 6 buttresses as anchorages;.the
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dome has a 3-way post-tensioning system. The foundation slab is reinforced with conventional reinforcing steel. The inside of the containment struc-ture is lined with a s, teel plate noted to be 1/4 in, thick in the drawing contained in the PSAR.
The design of the containment structure'Sor this facility is similar to that employed for the 71orida Power and Light Company's Tuckey Point Plant, and Consumers Power Company's Palisades Plant.
Three general types of steel prestressing tendon cystems have been under consideration by the applicant, namely those designated as Freyssinet, BBRV, and Stress Staci (S3ZE). The applicant has stated that the BBRV tendon l
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system is to be used in this facility. It is.noted that the tendons will be unbonded and protected against corrosion by. insertion of a petroleum based-product into the' tendon sheath.
e The reinforcing steel to be employed in the base slab of the containment structure will consist _ of steel bars conforming to ASTM designation A-432-65.
Rainforcing steel corresponding to ASTM A-15 or A-408 will~ be used in the cylinder wall and dome roof to control shrinkage and tensile cracks.
The liner plate will. conform to designation ASTM A-442 and ASTM A-20.
In areas of penetranion, steel conforming to designation ASTM A-516 Grade 60 or Grade 70 made to ASTM A-300 specifications will be employed.
The facility site will be founded on the Pliocene Laguna Formation which is underlain by an estimated 1500 to 2000 ft. of Tertiary or older sediments deposited on a basement complex of granitic or matamorphic rocks.
The geological explorations indicated that tha Laguna Formation is generally firm silt-stone, sand, gravel, and conglomerate. The compressional seismic f
velocities of the materials at and below the foundation leve1 of the facility are on the order of 2000 to 6000 fps.
The description of the foundation elevations for the reactor building, turbine building, etc., as presented in Section 2E of the PSAR indicates that the reactor containment foundation will oe located 35 ft. below the grade surface.
The finished grade elevation is noted to be at about elevation 165.
l The turbine building mat is noted to be located approximately 10 ft.
below the finished grade surface.
The nearest fault system, the '.*oothill Fault system, is 10 miles east.
of the site. The Hayward and San Andreas faults are 70 and 89 miles to the west, respectively. On the basis of the geological studies made and reported
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in Appendix 2C, it was concluded that no faulting of' the upper Tertiary and -
younger rocks has occurred at the site or in the surrounding area.
On the basis of the information. presented in the ?SAR, there appears to be little likelihood of liquef, action being a problem at this particular site and the foundation conditions appear to us to be adequate.
SOURCES OF STRESSES IN CONTAINMENT STRUCTURE A!D CLASS I COMPONENTS The containment vessel is to 'bo designed for the following loadings and conditions: dead load, live load, including snow and ice l'oads and equip-loads; prestress loadings; design accident ccmperature of 2860F and ment pressure of 59 psig; an air test pressure of 115 percent of the design pressura; an external pressure loading with a differencial of approximately 2 psi from outside to inside; wind loading corresponding to 90 miles per hour design wind at a reference 30 ft. above ground level; and earthquake loading as described next.
The seismic design is to be made for an Operating Basis Earthquake based
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upon a 0.13g maximum horizontal ground acceleration and a Design Basis Earth-quake based upon a 0.25g maximum horizontal ground acceleration.
The containment walls and liner are shielded by various types of barriers from impact from missiles which conceivably could have enough energy to penc rate them. The high-pressure reactor' cooling system equipment which could'be the source of missiles is screened either by the containment shield wall enclosing the reactor cooling loops, by the concrete operating floor, or by special missile shields to block any passage of missiles to the contain-cant valls.
The criteria controlling the design of piping and reactor internals f or seismic loading is presented in various places in the PSAR and scendments
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s but particularly in Section 3 of the PSAR, in Appendix 5A and in the answer to questions in Appendix SJ.
The loading combinations include consideration of design loads, pipa rupture loads, and earthquake loadings.
COMMENTS ON AD.0UACY 07 DESIGN Foundations and Dams The major facility structures are to be founded directly on competent soil and stone as' described earlier in this report. On the basis of the information presented in the PSAR and amendments, the foundation conditicns appear acceptable to us.
Appendices 2C and 2H of the PSAR describe the storage reservoir criteria.
It is noted there that an earch-fill dam will be employed to provide a reservoir of water for use by the facility, and that this reser-voir will.be located approxin.ately one mile cast of the plant. We are advised that this is a Class I ss.uctura bu: c.a t necessary f: safe shutdown.
The design safety factors against failure involving the maximum hypothetical carthquake is dispussed in Section 7 of Appendix 2C.
It is noted there that
the mini =um safety factor will be 1.0.
The discussion presented on page 2H-1 of Amendment No. 4 '.ndicates that flooding of the site arising from failure of the dam is unlikely to occur.' Further we understand that failure of 'the dam will not deprive.the plant of emergency cooling capability since this capability is provided by the two spray ponds; moreover, failure of the dam should not initiate a reactor accident.
Saismic Design Critoria We agree with the design criteria for an Operating Basis Earthquake of 0.13g maxi =um horisontal ground acceleration, with further provision that safe shutdown can be achieved for a Design Basis Earthquake of 0.250 These t
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's earthquake design values are in. agreement with those given by the U. S. Coast and Geodetic Survey (Ref. 2).
The response spectra for the Operating Basis Earthquake and Design- '
Basis Earthquake to be employed in the design are presented as Figs. SA-1 and SA-2 of Appendix SA of the pSAS. These spectra are scaled after chose presented in earlier publications by Dr. G. W. 'dousner, and we are in agree-ment with the spectra to be employed in the calculations.
The earthquake analyses will include the effects of vertical ground accel-e I
cration which in the answer to Question 5J.14, is noted will be taken as 1/2 of the horizontal ground acceleration.
In view of the location of this i
facility and the associated foundation conditions, it would be our recommen-dation that the ratio of vertical to horizontal ground acceleration be taken as 2/3 for a period less than 0.15 to 0.2 sec.
The applicant advises that for periods greater than 0.3 second the ratio will be reduced to one-half and for intermediate values of period, will be interpolated linearly to our i
i reco= mended value. We concur in this approach.
It is indi'cated in the answer to Question SJ 15 that the earthquake
'l loadings will be added linearly and directly as appropriate to the applicable loadings. We are in agreement with this criterion.
The method of dynam,1c, analysis to be, employed for the containment struc-ture is described in Section 5.1.5.6.
Although the method is not a standard rigorous one, it leads to conservative values of maximum shears and over-turning moments relative to the square root of the sum of the squares of the modal maxima "for these values computed in the standard way.
The damping associated with the response spectrum values used in the analysis is deter-mined from a weighted average of the damping values for separate components of motion including translation, base rotation, and relative translatory
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motion of the base, with weighting according to the combined modal accel-cration values for each of these components. This is frankly an approxi-macion.
The damping values used are summarized on page 5.1-32 and we are in gencral agreement with the values given for the structural and rocking damping. The translation damping of 3Q percent used for both carthquake designs is high, and it is our, understanding that this is applicable only to the containment structuro. This high amount of damping corresponds 4
essentially to direct translation of the base of the structure relative to the ground motion and probably results in little amplification.
This consideration of interaction of the containment foundation and structure and the surrounding ground also bears on the question of loadings to which the containment ctructure is subjected as a result of the earthquake ground motions. The information presented in the answers to Questions SJ.7.16 and 5J.16, indicates that reflected soil pressures will be used; this approach is conservative and we concur in this approach for this facility.
The loading combinations to be employed for the design of the contain-ment structure are given in Section 5.1.4 of the PSAR.
The earthquake and pipe rupture loadings are considered in the loading expressions involving
. factored loads, and the latter expressions appear appropriate for design of containment structure.
The loading expressions involving factored loads for structures other than the containment and the reservoir are presented in Section 3.13 of Appendix SA and appear appropriate to us.
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r The design of Class II structures is described in the answers to Ques-
.i tions 5J.6.5 and 5J.17.
It would be our recommendation that for those Class II items that are of special significance in terms of plant safety the design be made on the basis of about 78 percent of Zone 3 of the Uniform Building Code for structures employing damping values of 7 to 10 percent and a ducci-lity factor on the order of 3 to 6; for structures with lesser damping, on the order.of 2 to 5 percent, the static seismic design forces should be waitiplied by a factor of roughly 2.
Tha allowable stresses to be used with the design factored load expres-sions will be those corresponding to yield strength of the structure and will not involve significant deformation beyond the yield level.
In this way the load-deformation behavior of the structure is expected to be one of clastic, low-strain behavior in the sense that the loading does not exceed 1
the yield lavcl. We are in agreement with this criterion of design for the
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containment structure.
The approach outlined for handling principal concrete tension and
%e combined membrane tension and membrane shear appears acceptable to us.
The design of the liner and anchors is discussed in Section 5 of the PSAR. In the answers to Questions SJ.18, 5J.4, and 5J.20, it is noted that it has been decided that the' design will be st$ch as to make the critical elastic buckling strength below the actual yield strength of the plate rather than above it.
However, it is also noted that the liner strain criterion is such that strain in the liner is limited to 1/2 percent. On the assumpticn that the liner anchorage system is designed to be capable of handling the differential bending, shears, and axial loadings arising in adjacent panels as a result of liner buckling, we believe that the liner design can bc *.andicd satisfactorily.
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9 The general approach outlined for the prestressed design appears acceptable The applicant has centatively selected the IBKV prestr sing system to u.4.
. for the structure.
It is our opinion that any of the three systems considered by the applicant (Freyssinat, 33RV, and SEEE) could be acceptable provided that the applicant demonstrates satisfactorily that the system (including anchorages) will perform adequately under all types of loadings, including earthquake loading.
The design procedure for major penetrations provides methods for handling the membrane loadings as well as the secondary loadings, and appears satisfactory to us.
Class I Picine, Ecuincent, Vessels, and Reactor Internals The design criteria for Class I systems and components are presented in the answers.co Questions SJ.4.2 and SJ.7.27.
The loading expressions for Class I systems and equipment designs are
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listed on page SA-5 of Appendix SA.
We are in agreement with these loading
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The method o'f dynamic analysis for such systems is described on page 5J-4 of Appendix SJ, Altaough we agree in general with the approach presented, it,is our recomdendation that careful attention be given to.the boundary conditions, input loading levels, and the role of snubbers, dampers, and intermediate support points on the resulting design.
The design criteria presented, and the basis for estimatin'g the margins of safety, are based on stress criteria, which can exceed the yield icvel.
For the design of the piping, it is our recommendation that limiting strain criteria be employed also to insure that there is no loss of function or of ability for safe shutdown. The applicant has given such a limitation in Amendment No. 4 (page SA-5) and these criteria are acceptable to us.
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'10 The design criteria for reactor internals appear generally acceptable to us in carms of t'ge margins of safety that appear to exist from the design
, icvols given in the answer to Question 5J.4.2.
The seismic criteria for critical elements of control, instrumentation, bacteries, er.c., are discussed in the answer to Question SJ.24.
The design force icvels given there appear generally reasonable execpt for instrumentation, which is not given. The tilt tolerances noted are of some concern; the small values given in some cases seem questionable, as for exampic, the amount of permissibic tilt for a motor.
It would be our rer;mmendation that these criteria be examined in detail during the desiga phases, and also that similar design criteria 'for tha critical instrumentation be developed and checked during the design phase.
Tha design critoria for the cranes are given in the answers to Ques-tions 5J.3 and 5J.22; we are in agreement with the criteria as noted.
The matter of quality control, inspection and acceptance is discussed
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throughout the PSAR and amendments. The procedures outlined appear accept-able to us.
CONCLUDING COMMENTS On the basis of information presented in the PSAR and amendments and in keeping with the design goal of providing serviceable structures and components with a reserve of strength and ductility, we believe that the design outlined for the containment and other Class I structures and equip-ment and for Class II structures and components can provide an adequate margin of safety for seismic resistance.
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.s 11 R"7222'*02S 1.' "?reliminary Safety Analysis Raport U. Volumes I, II, III, IV, V (Including A:.and cacs 1, 2, 3 and 4)", Rancho Seco Nucict.: Cenor ting Scc: ion Unic No. 1, Sacrc=anto Municips'. U:111ty Dis'trict, 1967 and 1968.
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"12pc c: on :ha Saismicity of the Rancho Soco Nuclear Canarating Station, Unic No. 1 Site," U. S. Coast and Geodatic Survey, Rockville, Maryland, J..r.a 5,1968.
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