ML20235B687

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Draft Adequacy of Structural Criteria for Wh Zimmer Nuclear Power Station
ML20235B687
Person / Time
Site: 05000000, Zimmer
Issue date: 05/04/1971
From: Hall W, Hendron A, Newmark N
NATHAN M. NEWMARK CONSULTING ENGINEERING SERVICES
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References
FOIA-87-111 NUDOCS 8709240190
Download: ML20235B687 (8)
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Regulatory Fjle py;

. NATHAN M.'NEWMARK '"

CONSULTING ENGINEERING SERV!CES 1114 CIVIL ENGINEERING " BUILDING URBANA, ILLINOIS 61801 1

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REPORT TO AEC REGULATORY STAFF ADEQUACY OF STRUCTURAL CRITERI A FOR WILLI AM H. ZlMMER NUCLEAR POWER STATION CINCINNATI GAS AND ELECTRIC COMPANY, et al.

AEC Docket Nos. 50-358 and 5^  ::

by N. M. Neunark ,

W. J. Hall " OJ

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A. J. Hendron, Jr. s atC[pic Ch + -

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to 4 May 1971 8709240190 070921 PDR FOIA i-MENZB7-111 PDR

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ADEQUACY OF STRUCTURAL CRITERI A FOR WILLIM H. ZlHMER NUCLEAR POWER STATION by N. M. Newmark, W. J. H,all , and A. J. Hendron, J r.

INTRODUCTION This report concerns the adequacy of the containment structures and components of the William H. Zimmer Nuclear Power Station for which application for a construction permit has been made to the U.S. Atomic Energy Commission by the Cincinnati Gas and Electric Company, Columbus and Southern Ohio Electric Company, and the Dayton Power and Light Company. The facility is located 24 miles southeast of Cincinnati, Ohio on the Ohio side of the Ohio River, and approximately 1/2 mile north of Moscow, Ohio. This report is based on the infonnation and criteria presented in the Preliminary Safety Report (PSAR) and Amendments thereto as referenced herein. Additional Information has been obtained through discussions with the AEC Regulatory Staff. It is noted on I

pcge 1.1-1 that this P5AR covces cr.!y the first unit designated as the William H. Zimmer Nuclear Power Station, Unit 1.

DESCRIPTION OF FACILITY The William H. Zinsner Nuclear Power Station is described in the PS AR j as a single-cycle, forced-circulation, boiling water reactor producing steam )

i for direct u1e in the steam turbine-generator unit designed for a net electrical power output of about 807 MWe. l The primary containment, which houses the reactor vessel and other components, consists of a steel-lined prestressed concrete pressure suppression system of the over-and-under configuration. The drywell, in the fom of a cone, l

'l Is located directly above the suppression cha:nber. The suppression chamber, 1 1

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wMch is cylindrical, is separated from the drywell by a reinforced concrete slab which functions as the drywell floor. In the event of an accident, the drywell atmosphere is vented into the suppression chamber through a series of downcomer pipes penetrating the drywell floor. The drywell has a base diameter of approximately 80 ft. and a top diameter of about 30 f t.

The reactor building, which comprises the. secondary containment, will consist of poured-in-place reinforced concrete for the substructures and exterior walls of the building up to the refueling floor, and above this level the building structure will be steel-framed with insulated metal siding. The siding will have sealed joints, and entrance to the building will be through interlocked double doors.

LOADINGS AND SOURCES OF STRESSES The reactor containment structures will be desicned for th'e following I

loadings and conditions: dead loads; live loads; design temperatures and pressures j for the drywell and suppression chamber of +45 psig internal pressure and +2 psig external pressure, and temperatures of 290 F and 275 F for the drywell and j l

pressure suppression chambers, respectively; test pressures of 52 psig for the drywell and pressure suppression chambers; a maximum differential pressure of 25 psi app!!cd to the drywell side for the floor separating the drywell from the suppression chamber; a wind pr, essure ranging from 35 to 53 psf; and a tornado loading associated with a 300 mph horizontal peripheral tangential velocity with a transnational velocity of 60 mph, and an external pressure drop of 3 psi, with associated misslics.

F in addition, the design is to be made for a Design Basis Earthquake corresponding to a me.ximum horizontal ground acceleration of 0.20g (as noted in Acaendment 5) and for an Operating Basis Earthquake corresponding to a maximum horizontal ground acceleration of 0.10g. The vertical acceleration is to be taken as 2/3 of the horizontal value. Also, it is noted in the PSAR that the

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DBE. design will be for 0.139 at rock surface and the OBE design for 0.0659 at rock surf ace; comments on these design values are made later.

ADEQUACY OF DESIGN Foundation 2 Bedrock formations in the site area consist of limestone and shale, and are encountered at depths ranging from 83 to 88 f t. In the plant construction are a. Immediately above the bedrock for a depth of roughly 60 f t., the soll consists of medium dense fine to medium and fine to coarse sand with varying amounts of sil t and gravel, and occasional gravelly layers. The next 20 to 30 ft.

above that is characterized by dense to medlun dense and stiff interlayered silty fine sand, fine sand and clay, sit t and fine sand. The top layer, up to 12 f t, in thickness, consists of very stiff clay and fine sandy sit t with roots. The upper 30 to 35 ft. of soll is noted to be. recent alluvial deposits and the underlying sands are believed to be of glaclofluvial origin of Pleistocene age.

The nearest known fault is the Maysville Fault located about 30 miles southeast of the facility.

The answer to Question 2.5.4-1 of Amendment 4 Indicates that plie

. foundations will provide foundation support for all structures. This statement is evidently superseded by Amendment 5, where on page 2.5-81 (knendment 5) it is Indicated that major plant structures will be supported on mat foundations; the foundation elevations are given in Table 2.5-16. Foundation preparation will consist of densification of soils between foundation level and Elevation 450 by i

dewatering, excavating and recompacting existing soils. Although the general method described for construction of the foundations appears adequate, there is still one matter which remains to be resolved, namely certain details of the liquefaction analysis presented in the PSAR. We are in agreement with the water l

level used for the liquefaction analysis, namely Elevation 508. However, in

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evaluating the analysis it is' not known what densities of sand were used for t

-various depths, and the conclusions do not Indicate if the analysis is for the ,

natural soil profile or for the. compacted soll surrounding the reactor facility. l The applicant should clarify these points.

The service water pumphouse arrangement illustrated in the sketches in the PSAR appears satisfactory. However, we should like to know more about the material to be placed behind the sheet plie bulkhead and the pressures for which the sheet pile bulkheads are designed before coamenting finally on this aspect of the design.

The design configuration for the water conduits from the reactor building to the pumphouse has not been finalized. I t is understood that the design may be -

that of pipes supported on pile bents or alternatively in an excavated ditch.

In the infonnation presented in the PSAR it is noted that there are evidently locations below Elevation 465 where there are loose sands; thus, if the di tch were ,used, It would have to be deeper than Elevation-465. Further. detall s from tha applicant on the propcsed pipe design are required.

Connents on the river bank stability are noted in the answer to question A.3.1.1-5 (Anendment 4) and again on page 2.5-98 (Amendment 5) . It is indicated in the latter that analysis of the river bank stability will be performed for both static and dynamic loading conditions, and that consideration will be given to the effects of hydraulic fill placement, surche ge adjacent to the fill, and in situ and fill slope configuration. The analytical procedures that will be followed in evaluating the bank stability are evidently those described generally in the former reference; the appilcant should provide additional information about the details of the analysis procedures to be employed. Also, from the information provided, it is not possible to ascertain what effects bank stability could have on the safety of the plant.

5 Seismic design

'The earthquake design criteria and earthquake hazard were discussed in meetings with the AEC/DRS/DRL staff and representatives of NOS and USGS.

.it is our understanding that the hazard for which the plant is being designed is for a Design Basis Earthquake characterized by a maximun horizontal ground acceleration of 0.20g and for an Operating Basis Earthquake of 0.109 . The lower seismic values at bedrock as presented in the PSAR, namely 0.139 and 0.065g,

. appear to us to have little applicable significance insofar as design is concerned, in view of the essentially thin layer of overburden lying over the bedrock. On the assumption that. the design is carried out for a DBE of 0.20g and for an OBE of 0.109 for all Class I structures, we concur in the design values selected.

The response spectra for the OBE and DBE are presented in Figs. 2.5-22 and 2.5-23  !

(knendment 5) and we concur in use of the spectra presented there.  !

Moduli and soil damping values for the foundation materials are presented in Table 2.S-14. No specific values are given there for the compacted -

soils which will be employed under the facility structures. It would be our recommendation that a value not greater than 7 percent be taken for the soil damping values; it will be noted that this falls within the range of values given by the applicant in the table cited.

The methods of seismic analysis to be employed are described generally in Section 12.3 of the PSAR. This section, and answers to nany of the questions in Section 12, refer to Sargent r, Lundy Report SL-2690 (Ref. 2). The analytical procedures outlined there generally, and the damping values given for structural elements in Table 1 of Appendix A, are acceptable to us. There is one point, l

however, which needs clarification, and that refers to the discussion on page 9, j Section 2.2.5, of Report SL-2690 concerning the use of the four earthquakes with l

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a time-history analysis, and specification that the average of the exact spectra )

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i however, for all Classification Group B piping designed for the DBE, it is indicated that the piping is to remain functional by elther keeping the sum of the longitudinal stresses less than the minimum specific yield stress, or stresses above yield are allowed provided a plastic analysis' shows no loss of function.

The applicant should advise if there is a strain limit associated with the latter approach, and if so, its magnitude. -

No seismic analysis procedure for pressure vessels and heat exchangers were described in Appendix A. The appilcant should describe how these analyses 5

are handl ed.

1 In Section A.3-1 there is a statement that "In the seismic analysis category, the major participating analytical groups are the Engineer-Architect 3 cnd the NSSS supplier. Each group has developed 'Its own analytical techniques, procedures, and computer codes for seismic analysis". From this statement it is not clear whether the seismic design criteria given in Section A.3 are those of the NSSS or the Engineer-Architect; we had assumed that the procedures of the Engineer-Architect were described in Report SL-2690; and that the NSSS criteria f

were those being described in Section A.3. Clarification on this point is needed. i The general criteria described for buried piping (Section A.3.1.1.3)

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are acceptable with the exception that it is recommended that special design l j

precautions be taken at points where piping enters or leaves structures or other points of constraint.

Equipment (including reactor internals)

The equipment design criteria are presented in Appendix C and only a general description of the approach to be followed is presented there. A more detailed description is required if comment on the adequacy of the procedures are to be made. '

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- REFERENCES

" Preliminary Safety Analysis Report", ;Vol s.1-5, and Anendments 1, 4, 5 and

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6, . William H. Zimmer Nuclear Power Station, the Cincinnati Gas and Electric .;

Company, et al ., .1970 and 1971. l

2. " Procedures for the Seismic Analysis of Critical Nuclear Power Plant Structures,' Systems and Equipment" Report SL-2690, Sargent & Lundy Engineers, .

Chicago,13 Nov.1970 (Proprietary).  !

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