ML20127N586
| ML20127N586 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 01/07/1970 |
| From: | Hall W, Newmark N NATHAN M. NEWMARK CONSULTING ENGINEERING SERVICES |
| To: | US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML20127N578 | List: |
| References | |
| NUDOCS 9212010317 | |
| Download: ML20127N586 (13) | |
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- 4 REPORT TO AEC REGl'LATORY STAFT ADEQUACY OT STRUCTUPJJ. DESIGN FOR MONTICELLO UNIT NO. 1 Northern States Power Company AEC Docket No. 50-263 f
by N. M. Newmark W. J. Hall l.
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-1 0 4 ADEQUACY OF THE_ STRUCTURAL DESIGN FOR MONTICELLO UNIT NO. 1 Northern States Power Company -
by N. M. Newmark and W. J. Hall INTRODUCTION This report on Monticello Nuclear Generating Plant' Unit.1 was prepared on the basis of the following: (1) A review of the Final Safety Analysis Report (FSAR) and amendments thereto as submitted'by the Northern States Power Company, as listed at--the end of'this report; (a) a visit to the site by Dr. W. J. Hall on 17 September 1969; and (3) discussions of the facility with the AEC Regulatory Staff.
Monticello. Unit No. 1, as noted in Section 1 of the_PSAR, has been designed for a maximum output of 1670~ MWt _ (545 MWe- net) and consists of -
a single-cycle forced circulation, boiling water reactor _that_ produces s: cam for direct use in the steam turbine.- The_ facility is located 22 miles downstream'from St.. Cloud, Minnesota-and about three miles north-
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west of the village of Monticello, Minnesota, on the south bank of the Mississippi River.
The plant was designed for a Design Basis Earthquake of 0.12g maximum horizontal ground acceleration and for an Operating Basis Earth-quake of 0.06g maximum horizontal ground acceleration. The vertical carthquake excitation was assumed to occur simultaneously with the
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2 horizontal earthquake excitation for the design.of the Class I structures and equipnent, and was taken'as'two-thirds of the magnitude of the corresponding horizontal excitation.
The criteria applicable to the design of Monticello Unit'l were reviewed extensively by us at the construction permit stage; accordingly we have not included herein, unless there was some specific reason for h
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doing so, topics covered in our earlier report dated March 1967 (Ref. 2).
COMMENTS ON ADEQUACY OF DESIGN
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1 Foundations From study of the material presented in the FSAR and from diftve9fons 4
with the applicant and General Electric and Bechtel personnel at the site, it was cscertained that in accord with the criteria presented in the PSAR the clay was removed from underneath the reactor building, and the base of the reactor building was founded at elevation 888 on selected and compacted granular material,lwhich overlays hardpan and thence bedrock
[ (approx. elev. 860). The base of the turbine building is located at elevation 903 and there was no major amount of material removed and' s
replaced under the foundation of this building The-base:of:the intake structure is located at elevation 877; the foundation was placed on l-lean concrete which in turn rests on bedrock.
The stack is-constructed on a spread footing on original grade; L
L the footing is approximately 60 ft. across1 flats and is octagonally.
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'3 Building Separation _
f it was noted in discussions that there is a=one-inch' separation between the reactor building and turbine building;;this separation.is
.1 filled with a one-inch premolded filling. Studies reported by the applicant (p. 12-2.14 dated 10/17/69) suggest that racking between' the reactor building and turbine building should not occur. -
Dynamic Analyses of Structures and Piping ,
Two methods of dynamic analysis were employed for major structures c
and' are described by the applicant in ' general terms in the FSAR and-
. amendments thereto.
Time History -
It is noted in. Appendix'A to the FSAR that the-control room, off gas stack, reactor building, drywell, and reactor pressure vessel, were analyzed by the time history method. 'The solution:
_. in this case is based on a classical normal load approach for uncoupling-the equations of motion, and the time _ history of response _ for each mode is computed making use of the numerical integration of the Duhamel -
integral. The modal responses are combined at each instance'of time,
- and the maximum values are determined for use in the design process. .
It was indicated in Section 12 of the FSAR that the July.-21, 1952, Ni69 W Taft acceleration-time record was used for theftime-history .
analysis af ter scaling to the spec 1fied design level.- The response spectra for 2 percent damping for the time-history record:is_ compared P
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with the'Taf t normalized spectra given 'in the PSAR or' i ginally, and with Housner's 2 percent' spectrum, in Fig. 12-2-9. The comparison given there suggests that the time-history record of 10 to 12.sec.- duration -
leads to a reasonable and satisfrictory response spectra.
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From1t he.
discussion presented in Section 2.1.9 we are led to believe that this same length of record also was employed in the time-history analyses.
It is noted in Fig. 12-2-9 that there are some low-troughs in the spectra which fall below the recommended response spectra. The applicant-notes. '
on page 12-2-8 of the PSAR that: "When the time-history analysis was made, periods were_ examined and the corresponding spectra accelerations from both the time-history and response acceleration spectra were compared.
If substantial differences were-noted,:the model was modified-and a re-analysis was made. In this manner it was possible to avoid using spectral accelerations that were in a valley' of a time-history spectra." This approach is satisfactory if the'model finally used accurately represents the as-built structure. On=page 12-2.9 dated' 10/17/69 the applicant advises that re-examination of the analyses shows that there were no cases where the periods. fell in.the valleys of the spectra.
We believe that the time-history method-as used constituted a reasonably conservative ~ design approach.
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Response Spectrum - In Appendix A it is noted that the 20-inch suction header, the pressure suppression chamber, and the recirculation lines were analyred by the response-spectrum method of analysis. An amplified description of the response spectrum method, as used, is given in Amendment 23.
The Method Il modal response approach, as described, involves a calculation of the values of moment, shear, etc., for each mode and cc bining these by taking the square root of the sums of the squares of the modal response values. When several of the frequencies lie close together, however, even this approach may not be conservative and direct summation of the responses for those modes may be more appropriate.
Method I is described as a method by which the ine;rtia forces are computed separately for each mode, and the final inertial loading is taken as the combination (by the square root of the sum of the squares) of the modal values, with the moments and stresses being determined irca the resultant inertial loading so computed. We agree that this method is generally conservative, but not always so. However, the exceptions are not significant, and we accept the use of this method.
Method 11 is acceptable in all cases except where modal frequencies are nearly the same. In these cases , neither method is completely conservative, but they both are generally acceptable for the cases considered.
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6 The summary comparison _of stress presented'in Table 12-2-1B, applicable to Dresden Units 2 and 3, but stated as being typically ,
1 applicable to Monticello, suggest that stresses computed by Method I- j are generally larger, but not greatly so, than those computed by Method 11. In any event, the important point is that the stresses eniculated by either method are low for Dresden 2 and 3 and could be even lower for Monticello.
On the basis of the stress data supplied, we conclude that the design appears satisfactory for _ those items for which either. Method I l
as described or Method II was used, or for which a revised and accurate static analysis was used, as discussed below.-
1 Equivalent Static Loading Technique -The comments given above in this report relating to methods of analysis are applicable to piping as well as structural elements. It is understood from the FSAR that all piping, except for the recirculation lines, which were referenced 4
as being analyzed by the response spectrum: method, was analyzed by the method described in the FSAR beginning on.page 12-2.10, which is~
an equivalent static loading technique of-dynamic analysis. The
- discussion on page 12-2.9C (10/17/69) indicates that the 20 in..suctioni header also was analyzed by the latter technique, 1 The discussion on p. 12-12.11 (10/17/69). indicates that all.the piping summary data in Table 12-2-1C were obtained by the static analysis technique. Instrumentation piping was analyzed by a special procedure y
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7 specified by the seismic consultant to the applicant, as discussed briefly on pages 12-2.11 and 12-2,12, both dated 10/17/69. The 20 in.
suction line header tabulation indicates a satisfactory design for this item. ,
The discussion presented on page 12-2.10 et seg indicates that __
the piping system supports were located in a manner so as to be out of the " resonance range." In connection with this technique of analysis, the floor response spectra were employed in the analyses. From the discussion on pages 12-2.12 and 12-2.12A, both dated 10/17/69, it is noted that the motions at each of the individual support points were ,
considered in the ar.alysis through an averaging technique, and an amplification factor corresponding to the amplified acceleration of the building at the level of the piping supports. The inertial loads were applied in a manner to represent the most severe conditions of _
loading (for example reversed loadings on adjacent spans) and thereby lead to an analysis representative of actual possible loadings.
In the discussions carried on at the site visit, it was noted that the stack, which is a Class I structure, has an 18 in diameter main pipe connecting it with the reactor building. It is noted on p. 12-2.22A as revised 11/24/69 that this piping has been considered to have the same strains as in the surrounding compacted fill, corresponding to a maximum stress of 7,300 psi. This statement appears reasonable.
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Design Basis Earthquake Analyses - The applicant has confirmed on page 12-2.3 that all Class I items were checked during the design to be sure that they met the applicable criteria for the Design Basis Earthquake as well as the Operating Basis Earthquake.
Class I Structures Located within Class II Structures f
The design provisions made for Class 1 items located in Class II structures is presented on page 12-2.8, and page 12-2.18, and appears acceptable generally.
It should be noted that certain portions of the intake structure and certain of the pumps are denoted as Class I. For example, the RHR service water pumps are Class I and the piping connecting the RHR service pump with the reactor building would therefore also be Class I. It is to be noted that this particular piping runs through a concrete tunnel connecting the intake structure with the turbine building, and with
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proper design provisions, there is no reason to believe that this piping will be seriously imperiled by relative motion between the intake structure, the turbine building, and the reactor building.
Torus Assembly and Suction licader Of concern in this plant is the design of the suction header which is concentric with the torus-shape pressure suppression chamber. The material presented in Appendix A on this subject is supplemented by a discussion on p. 12-2.11 dated 11/24/69. It appears that a static analysis was made with stresses computed elastically. One of the
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9 critical points in the system is the pipe connection joining the suction header and torus, especially where it fastens to the torus.. It is.noted that this junction is reinforced, but it falls rougbly mid-way between- ,
two torus ring supports and thus can be subject to considerable motion and flexure. The stress at this point for the DBE-is indicated as 32,470 .
psi. The allowabic stress is quoted as 3.0 Sm or 52,500 psi.- However, l/ ,
, the yield point at 300*C is stated as being actually 33,750 psi, and that at-room temperature 38,000 psi. Hence the stresses are acceptable.
Flooding It is noted in the FSAR and during site inspection that there are a number of motor-pump units located on the lower levels of the reactor .
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building. We were advised that at least two of these sections of the building, which contain equipment which are critical for safe shutdown and containment will be isolated to prevent their flooding should there.
be any leakage from-the torus or other piping. We concur in this approach. ,
Reactor Internals n
The design criteria for reactor internals -was presented on pages, i
12-2-5 and 12-2-6 of the FSAR. The discussion refers to elastic and.
plastic strains," and the technique employed is referenced. Tlus greatest' strain at any point-in the reactor internal is stated as being B.64%,
which is not greater than 20% of' strain at the maximum stress in the stress-strain curve, and le an acceptable level. ,,
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, , . 4 10 Containment Design Stress Criteria On page 12-2-4 et_seqLof the FSAR reference is made to stress limits of "150% of AISC allowables for structural steel. -90%_of-yield stress for reinforcing steel. -85% of ultimate stress for concretc."
The limits as given are not comparable limits so far as~providing margins --
of safety but the applicant notes on p. 12-2.4 (10/17/69) that the actual design stress level does not exceed 0.70 fe, which is satisfactory.
Personnel Hatch During the site visit it was observed that the double; lock personnel hatch to the drywell was cantilevered some distance from the drywell and1 '
biological shield without any apparent. support except at its base. The applicant states on page-12-2.22A (10/17/69) that this aspect of the' -j design has been checked and been found to be satisfactory.
Thermal Loadings The tabulation of loading considerations that is given beginning ,
on page 12-2.3 indicates that thermal expansion loadings were considered in the design, and in all applicable cases,-_ including piping and equip-ment, the thermal loads were combined directly as appropriate. "
Seismic Criteria for Equipment Procurement -
The criteria given in the PSAR and FSAR for Class I. equipment
, indicate that the-procurement includes provision to meetLseismic requirements, uAs indicative of such procurement practices, the applicant has requested the seismic design. criteria from vendors
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'll for several itens, as for example the main ' steam isolation valves and cs tJie battery racks. From the discussions held ~at the site, and the discussion on p. 12-2.2 (10/17/69) there is reason to believe that these requirements have been, or will be, checked carefully by-the applicant.
Critical Instrumentation and Controls In Amendment 17 to the FSAR the applicant has described in some detail the procedures to.be employed for evaluation, and in some cases
, testing, of critical controls ano instrumentation to help demonstrate that these items can perform satisfactorily under postulated accident conditions. We believe the criteria given therein to be adequate in evaluating these items of control and instrumentation at:this time inso-far as it is presently possible to do so.
Stack i
It is to be noted in this case the main vent stack has been located at a distance from the facility which would preclude its-striking the facility should it topple. *
SUMMARY
COMMENTS 6-After a review of the FSAR, and amendments thereto and the other J
. material made available to us through discussions, we believe that the D
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12 design .is reasonably adequate in terms of provision for.' safe' shutdown for a Design Basis Earthquake of 0.12g maximum horizontal ground ,
acceleration and to withstand otherwise_the effects of.an earthquake >
of half this magnitude.
REFEPINCES
- 1. " Final Safety Analysis Report, Vols. I through VII, and Amendments .10 11, 12, 14, 15, 16, 17, and 21, 22, 23, 24, Monticello Nuclear.
Generating Plant, Monticello, Minnesota, Unit 1, Northern States Power Company," 1968 and 1969.
- 2. " Adequacy of the Structural Criteria for the Monticello Nuclear Generating Plant Unit 1," prepared by N. M. Newmark and.'W. J. Hall, under AEC Contract No. AT(49-5)-2667, March'1967.
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