ML19210A961

Adequacy of Structural Criteria for TMI-1, Draft

ML19210A961
Person / Time
Site:	Three Mile Island
Issue date:	10/31/1967
From:	Hall W, Newmark N ILLINOIS, UNIV. OF, URBANA, IL, NATHAN M. NEWMARK CONSULTING ENGINEERING SERVICES
To:
Shared Package
ML19210A959	List: ML19210A958 ML19210A961
References
NUDOCS 7911010670
Download: ML19210A961 (10)

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sNATHAN M. N E h ,.i A R K CONSULTING ENGINEERING SERVICES 1114 CIVIL ENGINEERING DUILDING URDANA. ILLINOIS 61801 DRAFT Report to AEC Regulatory Staf f ADEQUACY OF THE STRUCTURAL CRITERIA FOR THREE MILE ISLAND NUCLEAR STAil0N UNIT I Metropolitan Edison Company (Docket 50-289) by N. M. Newmark and W. J. Hall October 1967 3g ,s q qe q lsu/ LUU 7911010 ( 7 0 3495.

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ADECUACY OF THE STRUCTURAL CR;TERI A FOR THREE MILE ISLAND NUCLEAR STATION UNIT 1 by N. M. Newmark and W. J. Hall INTRODUCTION This report is concerned with the adequacy of the containment structures and components for Three Mile Island Nuclear Station Unit I for which application for a construction permit has been made to the U. S. Atomic Energy Commission by the Metropolitan Edison Company. The facility is located on Three Mile Island near the east shore of the Susquehanna River in Dauphin County, Pennsylvania.

Speci fically this report is concerned with the design criteria that determine the ability of the containment system, piping, and other critical components to wi thstend a design earthquake of' 0.06g maximum transient horizontal ground acceleration sinultaneously with the other applicable loads forming the basis of the design. The facility also is to be designed to withstand a maximum earthquake of 0.129 horicontal ground acceleration to the extent of insuring safe shutdown.

This report is based on information and criteria set forth in the Preliminary Safety Analysis Reports (PSAR) and the supplements thereto as listed at the end of this report. Also, we have participated in discussions wi th tne AEJ regulatory staf f concerning the design of this unit.

DESCRIPTION OF FACILITY Three Mile Islar.d Unit 1 is described in the PSAR as being a com-plete nuclear power unit, licensed for coeration at power levels of 2452 Mwt (845 MWe) . The nuclear steam supply system is to consist of a pressuriced water reactor to be supplied by the Babcock and Wilcox Company, quite similar 1589 28*

to other units now under construction such as Oconee Nuclear Station Units 1 and 2, Turkey Point Units N . 3 and 4, and Indian Point Unit flo . 2.

The containment building, consisting of a prestressed, post-tension concrete reactor building, is similar in design to that employed for Turkey Point, Palisades and Point Beach. The containment building has an inside diameter of about 130 feet, is about 187 feet high, with a cylindrical wall thickness of 3 feet, and the dome thickness is 3h feet. The foundation mat will be approximately 9-foot thick with a 2-foot tnick concrete slab above the bottom liner plate. The liner plate thickness will be 3/3 inches for the cylinder and dome and 1/4 inch for the base. Unlike most other designs of this type, wi thin the containment building the two steam generators are enclosed in semicircular containment barriers

• i.e. there is a rather substantial concrete wall between the steam generator and the wall of the containment building.

The geology report indicates that the bed rock surface at the site is essentially flat and consists of Gettysburg shale. The bed rock at the site is overlain by a fluvial sand and gravel of about 20 ft. thickness containing varying amounts of silt, clay, and occasional lenses of clear sand. In the vicinity of the proposed plant site the death of soil over bedrock is about 20 feet. Section 2.1 of the PSAR indi:ates that the con-tainment structure will be founded on the bedrock. H owe ve r , in examining the elevations in Section 1 of the PSAR it appears that not all structures will be founded on the si' ale but many of them will be founded a t grade and at other elevations. It was reported that there i s no evi dence of faul ting at the site.

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The reactor containment building is to be designed for the following loads: deadicad; liveload; an internal pressure of 55 psi, a test

w pressure of 63.3 psi', an external atmospheric pressure 2.5 psi greater than that of the internal pressure; an external vacuum of 3 psig; a tangential wind velocity of 300 mph corresponding to a tornado; normal wind loading correspond-ing to 100 year frequency wind as determined from ASCE Paper No. 3269 (wind Forces on Structures), an internal temperature under operating conditions of Il0'F; an accident temperature of 281*F; and prestressing loads, in addition, Section 5.1.2.3.4 indicates that the building design will take into account any buoyancy that may arise from flood conditions.

The reactor containment building will be designed for a naximun horizontal component of ground acceleration of 0.06g and for a maximum ground acceleration of 0.129 for no loss of function.

The discussion on page SA-1 Indicates that the criti;al Class I piping and internals will be cesigned for the same earthquake loadings jus t noted. A similar statement for the internals is given on page 3-53 of the PSAR. However, in the case of the piping and internals we do not have an indication of how the loads will be combined,'as will be discussed later.

Class ll components are to be designed for a maximum ground acceleration of 0.06g in accordance wi th the Uni form Building Code pro-cedures.

COMMENTS ON ADECUACY OF DESIGN Sei smic Desien Cri teria -- We agree wi th the ' approach involving a basic design for a design earthquake of 0.06g horizontal ground acceleration wi th the provision that safe shutdown can be achieved for a maximum earthquake of 0.12 9. These earthquake design values are in agree-ment with those given by the U. S. Coast and Geodetic Survey (Ref 3).

The vertical acceleration is noted to be taken as 2/3 of the hortcontal component, and we concur with this apcroach.

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. _4 The response spectra for the design earthquake were originally presented in Figures 1 and 2 of Appendix 28 of the PSAR and corresponded to the 1957 Golden Gate earthquake. Subsequently, the applicant notes in answer to Question 3.1 of Supplement I that the revised acceleration response spectra las presented in Figure 3.1-1) reflect the greater response at lower frequencies based upon the 1940 El Centro spectra. We have verified that the spectrum shown in Figure 3.1-1 corresponds roughly to the envelope of the spectra corresponding to the 1957 Golden Gate spectra and the 1940 El Centro spectra. It is difficult to interpret and check such spectra when plotted to an arithmetic scale. On the assumption that the envelope as noted,and not the individual earthquakes separately, will be employed in the design, we concur in the spectra as amended.

The method of analysis for handling the dynamic excitation is described on page SA-3 with revisions as described in Appendix SC. The design approach noted there appears acceptable to us.

However, further comment is given in answer to Question 3.1 wherein it is noted "for other than a modal analysis the mathematical models will be subjected to the ground motion described as acceleration as a function of time. The input ground motion shall individually be the 1957 Golden Gate Park, San Francisco earthquake, and the 1940 El Centro earthquake, both normalized to 0.069 for the design earthquake and 0.129 for the maximum probable earthquake. The design will sa ti s f y both accelograms. The design approach involving a time history of ground motion is acceptable to us only i f the combined motions of the two separate earthquakes are con-sidered, or as long as the design corresponds to the envelope of the two spectra noted. for any givcn value of damping.

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, The damping values to be employed in the design are listed on page SA-3 of the PSAR. It is noted in answer to question 3.2 of Supplement I that the damping values will be applied to the maximun carthquake as well as to the design earthquake. The damping values given are acceptable to us with the restriction that, should any concrete structures above ground be employed with a 5 percent damping value as noted, the cracking associated with this high damping value must be taken into account if it involves any equipment or other items associated with safe shutdown, it is noted that the damping for the reactor building is 2 percent which is acceptable to us.

The statement on page SA-2 of the PSAR indicates that the primary steady state stresses, when combined with the seismic stress and occurring simultaneously, shall be limited so that the function of the component system of structures will not be impaired in a manner that might prevent a safe and orderly shutdown. This statement indicates that the seismic stresses will be combined with the prim.ary stresses. A further elaboration of this point is given in answer to Question 7.2-1 where it is noted that the vertical and horizontal frequencies will be added. This latter statement is unacceptable for the matter here is one of adding appropriate stresses,not frequencies, arising from the seismic loadings wi th the stresses arising from deadload-liveload, etc. which occur at a particular point. These stresses are to be added directly and linearly in all cases. This matter requires further clarification on the part of the applicant.

With regard to the design of the reactor containment building. the combination of loadings and load factors presented on page 53-2 (revised 10-2-67) tre acceptable to us.

The matter of combination of loadings involving the dead load, pressure, and temperature loading arising from accident, and earthquake 4

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." loading receives further attention in answer to Cuestion 3.3-1 of Supplement 1.

The answer that is given there is not clear insofar as compounding of loadings is concerned for the design of components other than the reactor buildinc. One can infer generally from the discussion of pages 5A-3 and 5A-4 that all applicable loadings will be combined, but the s ta tements are not clear. It is our recommen-dation that applicable loading combinations be applied for piping, rea:ter internals and other Class i equipment in the same manner as for the reactor containment building, and that clarification of the statement given in answer to Question 3.3 must be provided by the applicant.

A discussion of the design of reactor building cranes is presented in answer to Question 3.2 of Supplement 1. The design criteria are still not clear even wi th reference to Appendix 5A, for in addition to stress design, the problem with cranes is to insure that ti.Jy cannot be displaced from the tra:k and in doing so damage equipment which would be required for safe shut-down. Further clarification on the crane design problem is necessary.

With regard to the overall design of the reactor containment build-ing, it is noted on page 58-2 of the PSAR that the load deformation behavior of the structure is one of elastic low-strain response. Furthermore, it is noted that the strain in the steel liner plate will not exceed one-hal f percent strain. These criteria are acceptable to us.

The design of the penetration is discussed on page 58-5 of the PSAR and further elaboration is given in answer to Question 7.3 of Supplenent 1, it is noted therein that a finite element solution will be emoloyed for the design and further that the equipment and personnel accesses will be reinforced to wi thstand comouted stress concentrations and to insure approximate strain compatibility within the shell, This approach is acceptable to us.

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The PSAR indicates that the containment building will be founded on bedrock. Many of the other buildings appear to be founded on soil at grade and this assumption is reinforced by the discussion given on cage 3.2-2 of Supplement I wherein there is a discussion of relative motions between the various buildings. The discussion there indicates that the alloaable stress values corresponding to the relative motions will be based on code values and held lower than normal. The exact meaning of this is not clear and it would be desirable to have some indication of the amounts of ductili ty that would be provided rather.than the stress limits, for this is primarily a deformation problem.

We find little discussion of the manner of ;arrying the shear arising from earthquake loadings in the structural reactor containment build-ing. Some comment is given on this matter on page SC-4 of the PSAR. It would be our recommendation that net principal tension not be permitted on a section which is required to carry shear; however, we are willing to permit a net principal tensile stress of 3s f' excluding bending or flexural stress and c

6s f' when local bending arising from ther.al loads is included.

C Tne answer to question 7.10 of Supplement I and Apoendix 50 of the PSAR indicates that Cadweld splices will be employed for all bars of sizes larger than no. II, We are in agreement wi th this aporoach.

Corrosion Prote: tion -- Section 5.1.2.a of the PSAR indicates that corrosion protection will be provided to protect the liner plate, the tendon steel casings, and all reinforcing bars below grade. We are in agreement with this approach.

Long Term Surveillance is discussed in answer to Ouestion 3.7 of

, Supclement 1. It is incicated there tnat there will be little need for long term surveillan:e other than that of moni toring crack;and dimens ional

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. changes of the structure. We concur in this latter approach but believe that it would be desirable to provide for periodic tendon inspection at the anchorages in addition to a systematic monitoring of structural dimensional changes, crack surveillance program, etc.

Picino and Internals -- S ome c ommen t s on t he p i p i ng de s i g n is con-tained in an>wer to Question 3.2 of Supplement 1. The discussion tnere indicates trat for the design earthquake the stresses will be held within allowable working stress limits as set forth in the Power Piping Code. For the no-loss-of-function earthquake the stresses are to be limited so that they do not exceed the material yield stress and so that the function of the connected component system will not be impaired and that a safe orderly shutdown of the plant can be made. This approach appears acceptable to us as long as the design includes a combination of all the applicable loads as noted earlier in our report, a point which remains to be clarified by the applicant. There is no specific comment in this connection with regard to the reactor internals and this needs to be provided by the applicant.

Intake Structure and Emercency Water for Shutdown Coolinc -- A discussion of this problem occurs in answer to Ques tion 2.2 of Supplemen t I and it appears that norral s tream channel flav will be suf ficient for handling shutdown in the event of failure of the New Haven Dam. It is indicated there that the intake structure will be constructed at an elevation to take water from the botto,of the river and to maintain mini,al sub.nergence from the intame ,amc c'. al' time 2 Na aentica ia r.a ce cf the proble c' .;

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an oGcJr at the Site Du! it I- A U U"9 3 Inat OrOVI;E9F 4 i'I be "a ,e 'or harjling ice 'oaiings, 10_NCLU 510N S On the cas.s of the information presented, and in seepin e .i :a the 1509 208

, design goal of providing serviceable structures and coiponents with a reserve

. of strength and ductility, we believe the design criteria outlined for the con-tainment systen and other Type I structures and equiprents can provide an adequa*.e margin of safety for seismic resistance if the topics listed below are resolved.

in reaching this conclusion we assune that the applicant will give further consideration to the matters of compounding the horizontal and vertical carth-quake stresses with other applicable stresses, loading criteria in terns of combinations of loads required for piping and internals, stability of cranes, relative motions between the buildings and the piping connections thereto. the stress criteria for shear, the design criteria for piping and internals, and ice loadings.

REFERENCES

1. " Preliminary Safety Analysis Report -- Volumes 1, 2. and 3," Three Mile Island Nuclear S tation Uni t 1, Metropolitan Edison Company, 1967.

2. " Preliminary Safety Analysis Report -- Volume 4 (Suppl . No. 1)," Three t Mile l> land Nuclear Station Unit 1, Metropolitan Edison Company, 1967 3 " Report on the Seismicity of the Three Mile Island Nuclear Station Unit 1,"

U. S. Coast and Geodetic Survey, Rockville, Maryland, V