ML20136D658
| ML20136D658 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 10/31/1985 |
| From: | BECHTEL GROUP, INC. |
| To: | |
| Shared Package | |
| ML20136D655 | List: |
| References | |
| 0291519, 291519, NUDOCS 8601060234 | |
| Download: ML20136D658 (98) | |
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1 i
l CONFIR11ATORY REACTOR BUILDING BASEf1AT ANALYSIS 4
FOR
, ilOPE CREEK GENERATING STATION -
PUBLIC SERVICE ELECTRIC AND GAS ODt1PANY V
i 1
i i
Bechtel Western Power Division Job 10855 San Francisco Area Of fice Revision 1 October 198 5
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CONFIRMATORY REACTOR BUILDING BASEMAT ANALYSIS Table of Contents ,
Page
1.0 INTRODUCTION
1 1.1 Objective s 1 1.2 Scope of Work 2 1.3 Summary and Conclusions 2 2.0 DESIGN BASIS 4 s 2.1 Loads 4 2.2 Load Combinations 5 2.3 Material Properties 5 2.4 Soil Spring 6 3.0 Analysis Procedure 6 4.0 Parametric Study 7 5.0 Finite Element Basemat and Superstructure Model 7 6.0 Analysis Performed and Results 8 7 . 0. Evaluation of Basemat Design 9 AP PENDICES : A - References A-1 I
B - Tables . B-1 l C - Figures C-1 l
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l LIST OF TABLES l'~ ( APPENDIX B)
Table Page 5-1 Summary of Dead Load (DL) , Operating Live. B-1 Load ( Lo) , a nd Live Load ( L) - Reactor Building.
5-2 Summary of Dead I. cad ( DL) , Ope ra ting Live B-2 Lo a d ( Lo ) , a nd L ive Lo a d ( L ) - Rad wa s t e Area.
5 -3 Horizontal North-South OBE Seimic Nodal B-3 Lo ad . ,
5-4 Horizontal East-West OBE Seisnic Nodal Load B-4 5-5 Added Vertical Nodal Loads Due to Horizontal B-5 -
North-South OBE Earthquake-Reactor Building 6-1 Loads and Load Combir ations B- 6 7-1 Summary of Reactor Building Basemat Design B-8 Evaluation.
( 7-2 Maximum Reinforcement and Concrete Stresses B-9 fran CECAP Analysi s.
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NW 7 'd5 f.2 91510 LIST OF FIGtRES
( APPENDIX C)
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Ficure Page 2-1 General Plant Layout C-1 2-2 Basemat Concrete Pour Sequence C-2 4-1 3 Layer Model for Paramtric Study C-3 4-2 4 Layer Model for Parmetric Study C-4 l 4-3 5 Layer Model for Parametric Study C-5 4-4 Plan View of Basement Model for Parametric C-6 Study .
4-5 Transverse Shear Stress Distribution C-7 ~-
Along Basemat Depth, at Centroid of Brick l Elemen t S tack, No. 1 4-6 Transverse Shear Stress Distribution Along C-8 Basemat Depth, at Centroid of Brick Element S tock No. 2
, 4-7 Transverse Shear Stress Distribution Along C-9 Basemat 4-8 Normal Stress Distribution Along Basemat C-10 Depth, at Centroid of Brick Element Block No. 1 4-9 Normal Stress Distribution Along Basemat C-ll 5-1 An Isometric View of the Basemat C-12 5-2 An Isometric View of the Superstructure C-13 5-3 N-S Section of Basemat and Supe rstructure C-14 Model 5-4 E-U Section of Basema t and Supe rs tructure C-15 Model 5-5 Arrangement of Plate Element - Plan View C-16 5-6 Developed Elevation of Cylinder - South-East C-17 Quad rant 5-7 Developed Eleva tion of Cylinder - North-East C-18 Ouadrant F90/3 iii e , . , m.--.. - _,.-- - - - - - - -
Figure Paa 5-8 Developed Elevation of Cylinder - North-West C-19
(. Quadrant 5-9 Developed Elevation of Cylinder - South-West C-2 0 Quadrant 5-10 Section A-A C-21 5-11 Section B-B C-23 5-12 Section C-C C-2 5 5-13 Section D-D C-26 5-14 Section E-E C-27 5-15(a)' Section G-G C-27 5-15 ( b ) Section H-H C-29 _.
5-16(a) Section F-F , C-3 0 4
5-16 ( b ) Section I-I C-31
! 5-17 Sections L-L, M-M, K-K C-3 2 5-18 Section N-N C-3 3 5-19 Section 0-0 C-3 4 5-20 Sections.P-P, 0-0 C-3 5 5-21 Sections R-R, W-W C-36 5-22 Section S-S C-3 7 5-23 Section T-T C-38 5-24 Beam Elements C-3 9 5-25 Reactor Building Basemat Model - Plan View C-40 5-26 Designation of Walls C-41 5-27 Designation of Walls and Columns-Radwaste C-4 2 Area 6 -1 Resultant Transverse Shear (V y) Con tour C-4 3 Under Dead Load
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Figure Page 6-2 Resultant Transverse Shear (v g) Con tour 0-44 Under Dead Load 6-3 Resultant Moment (My ) Contour Under Dead C-4 5 Load.
6-4 Resultant Moment (Mz) Contour Under Dead C-4 6 Load.
6-5 Resultant Moment (Mt) Contour Under Dead C-4 7 Load.
6-6 Elements with Resultant Transverse Shear (V) C-4 8 4
Exceeding 100 kip /f t.
6-7 Elements with Resultant Moment (My) Exceed- C-4 9 -
ing 1000 kip ,f t/f t.
6-8 Elements with Resultant Moment (Mz) Ex ceed- C-50 ing 1000 kip-f t/f t.
6-9 Elements with Resultant Twisting Moment (Mt) C-51 Exceeding 500 kip-f t/f t.
6-10 Vertical Soil Pressure Contour (KSF) Due to C-52 Dead Load.
6-11 Vertical Basemat Deflection Due to Dead Ioad C-53 (Looking North) 6-12 Vertical Basemat Deflection Due to Dead Load C-54 (Looking East) 6-13 Vertical Deflection Contour - Load Case 14 C-55 l 6-14 Vertical Deflection Contour - Load Case 15 C-56
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6-15 Vertical Deflection Contour - Load Case 16 C-57 6-16 Vertical Deflection Contour - Load Case 17 C-58
- 6-17 Vertical Deflection Contour - Load Case 18 C-59 6-18 Vertical Deflection Contour - Load Case 19 C-60 7-1 Basemat Potential Shear Failure Modes C-61 7-2 Basemat Vertical Construction Joints C- 6 2
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l.0 INTRODUCTION
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1.1 Objectives The Reactor Building basema t wa s one of the specific ele- l ments selected by the Independent Design Verification '
Program (IDVP) team to determine whether concrete struct-ures of the Hope Creek Generating Station were designed l i
according to applicable design codes and FSAR conmitments.
The Reactor Building basemat was selected because it is a major structural camponent supporting the building and is subjected to various unique loadings. The I DVP team found that the design procedure for the concrete structures is in agreement with the applicable design codes, design criteria, and FSAR conmi tments. It was also concluded that the design of the basemat addressed all major loads.
However, the IDVP team had the following conments (Ref. 1-1 ):
- 1. The finite element mode' ling of the basemat utilized -
a coarse mesh that the team believed did not pennit verification 'of the adequacy of the mat (OR 7, Ref.
1-2).
- 2. The effect of twisting moments on the design of re-inforcing was considered during an independent review, however, documentation of this review is no
(' longer available.
- 3. A capacity reduction factor of 9 = 0.9 was used.
However, documentation of the adequacy of this value was not found in the design calculations.
- 4. The shear design of vertical construction joints was not explicitly addressed (OR 4 0, Ref. 1-3).
- 5. The team requested that PSEsG and Bechtel confirm the following statenents:
- a. Thermal loads, torus uplif t load , and seismic inertia loads due to containment flooding were " ~
not considered to be governing in the design of the basema t, a nd
- b. Flexural and shear stresses in the mat were small ~#-~
and the minimum reinforcement requirements of ACI 318-71 provided suf ficient margin for all design loads. Also , c omput ed mat s tr es ses we r e,
, in general, below the rupture modulus of concrete.
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The objectives of this confirmatory Reactor Building base-
_[ mat analysis are to address all the IDVP team's review
\ comments and to reinforce the adequacy of the basemat design.
1.2 Scope of Work This report summarizes the results of the confirmatory Reactor Building basemat analysis performed by Bechtel for the Hope Creek Generating Station. The work per-formed in this analysis is presented as follows:
o Section 2 presents the design basis of the analysis, o Section 3 describes the analytical procedure g used in the analysis, o Section 4 discusses the parametric study performed to determine the number of brick elements to be used through the thickness of the basemat model in order to provide element stresses which are suffi-ciently accurate to obtain the appropriate bending moment in the basemat.
o Section 5 describes the development of the finite element basemat and superstructure model and the loads applied to the model such that the resulting
( shear and moment at the basemat level envelop the gross shear and moment obtained from the seismic soil-structure interaction analysis of the Reactor Building. The application of seismic overturning moments to the basemat model is consistent with the method assumed for the design of the shear walls.
7
< o Section 6 discusses the static analyses performed i for the basemat and presents the analytical results l
which were used to confirm the adequacy of the basemat design.
o Section 7 presents the evaluation performed to verify
- the adequacy of the basemat design. All IDVP team's comments as described in Section 1.1, are addressed.
1.3 Summary and Conclusions The Reactor Building basemat confirmatory analysis was performed using a finite element model with element sizes finer than those used in the original basemat analysis.
Static analyses of the basemat were performed for three
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critical load conbinations to reconfirm the design ade-
- g quacy of the critical areas of the ba sema t under criti-A cal (vertical) lo adi ng s. The application of the seismic overturning moments to the Reactor Building basema t model is consistent with the method used for the design of shear walls. The results of this confirmatory analysis can be summarized as follows
- 1. The finite element model with five brick elements through the thickness of the basement will provide element stresses which are suf ficiently accurate to obtain appropriate bending noments in the base-mat.
- 2. The reinforcement pr ovided is adequate for all major loadings including the ef fects of the twis ti ng moments. The inclusion of twi s ting mome n ts is documented in Section 7.5.
- 3. The capacity reduction factor is a f unction of the ..
compressive force acting on each section. Th e min imum '
value used in the evaluation of the conbined moment and axial force is 0.84 (as allowed by ACI 318-71).
The - basemat design is found to be adequa te.
- 4. The shear strength at cross-sections containing a construction joint is found to be adequate to resist the induced transverse shear stress.
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- 5. The rmal loads, torus uplif t loads, and seimaic inertia ,
loads due to containnent flooding can be adequately accanmoda t ed in the ba sema t de sig n.
- 6. The computed basemat flexural and shear stresses in concrete are generally small. All conputed basema t tensile stresses are below the modulus of rup ture for concrete, except in a localized area close to A the drywell pedestal. The provided reinforcement di adequately resists these tensile stresses.
Based on the above , it is concluded that the original de-sig n of the Reactor Building ba sema t is adequa te. The basemat design addresses all major loads and meet all the i applicable design code, design criteria, and FSAR commit-ments.
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N07 7 as r.2 91519 2.0 Design Basis
(. The relative sizes and location of the major plant structures are shown in Figure 2-1. The overall dimensions _ of the power block are approximately 500 f t by 600 f t. The foundation for the power block area consists of fiiS basemats separated by 2 inch seismic joints. The Reactor Building ba sema t ( 19 2 '-6 " by 312 '-0") rests on two six-inch layers of wearing concrete sepa-rated by a 0.25" thickness of waterproofing membrance. The wearing concrete is founded on a compacted soil backfill which in turn overlies the Vincentown Formation. The 14-foot basemat was poured in a series of blocks in two seven-foot thick lif ts of concrete. Both horizontal and vertical construction joints were made in accordance with Hope Creek Specification 10855-C-103(Q) (Ref. 2-1) and ACI 318-71.
All construction joints were cleaned by sandblasting. The s ur-face of the concrete was then washed thoroughly to remove all loose material. The existing surf ace was thoroughly wetted before placing new concrete. The sizes of these concrete blocks -
range from 56' by 64' to 9 6. 25 ' by 104 ' . Figure 2-2 shows the plan view of the basemat with the pour sequence of the concrete blocks in both the upper and the lower lif ts.
The following is the design basis of the analysis:
2.1 Loads The loading conditions and applicable load f actors for the basemat design are enumerated in the General Civil-
, Structural Design Criteria, 10 855-D2.1 (Ref. 2-2). The i
loads applied at the top of the Reactor Building basemat (Elev. 54 '-0 ") consist of the following s
- a. Dead load (DL) from building walls, equipment and components on Elev. 54'-0", including t he toru s, and buoyancy loads.
- b. Ope rating live load ( Lo) from building walls, columns and the Elev. 54 '-0" floor,
- c. Live load ( L) from building walls, columns and the Elev. 5 4 '-0 " floor.
- d. Post accident containment flooding loads (F).
- e. Hori zontal North-S outh, Eas t-West and Vert ical seismic loads (OBE) from building walls, bat zma t, and torus (Eo, Ref. 2-3 ) , a nd
- f. Safety Relief Valve ( S RV ) d i sc h a rge loads (Ref. 2-4 )
j are included as part of the ope ra ti ng live load . The direction of the SRV discharge loads can either be up-( ward (uplif t) or downward.
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2.2 Load Combinations k Based on a review of original bassnat analysis (Ref. 2-5) results and discussions with IDVP team during review me e ti ng s , the following three load combinations were considered in the confirmatory basemat analysis.
(1) U = 1. 4 D L + 1. 7 Lo + 1. 9 Eo (2) U = DL + Lo + Eo + F (3) U = 1. 4DL + 1.7L + F As discussed in the Appendix 3D of the FSAR, Load Combina-tions (1) and (3 ) governs the shear and the flexural rein-forcement requirements respectively for the basemat. Load combination (2) was considered in order to address the IDVP team's comments that the to ru s upli f t load i ng a nd the seis-mic inertia forces due to the post accident containment flooding were not taken into account in de te nni ning the design moment and shear in the original basemat analysis.
As stated in Section 2.1, the SRV discharge loads are included as part of the operating live lo ad . Also , the direction of the SRV discharge loads can either be down-wa rd or u pwa rd . However, for the Load Combination (2) in which both the operating live load and post accident con-ta inment flooding load are cons idered , the SRV discharge loads are excluded from Lo. Post accident containment g' flooding loads and SRV discharge loads will not occur at the same t ime .
2.3 flaterial Properties The concrete used has the following properties:
Specified 11inimum Compressive Strength, fc = 4.0 ksi Modulus of Elasticity, E c = 5.47 x 10 5 ksf Weight of Concrete, w= .15 kcf Po i s so n 's R a t io , y = .17 The specified minimum yield strength of non-prestressed rei n forcene n t , fy, is assumed to be 60 ksi.
The minimum in-situ concrete strength obtained f ran con-pressive cylinder tests shows that the actual concrete strength is a minimum of 4,886 psi at 97.7% confidence level. The typical in-situ average yield strength of the
/f(
reinf orceme n t is approxima tely 13% higher than the specified minimum yield strength.
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2.4 Soil Springs The spring constants for the horizontal North-South, East-West and Vertical soil springs used in the analysis are 61.1, 58.1 and 210 kips / f t3 respe ct ively. These values are computed assuming the foundation soil is a uniform linear visco-olastic half space; and using the weighted average of the strain compatible (final iterated) shear modulus of the foundation soil obtained from the seismic soil-structure interaction analysis of the Reactor Building (Ref. 2-6).
- 3.0 Analysis Procedure The Reactor Building foundation mat was analyzed using the Bechtel Structural Analysis Program ( BS AP, Ref. 3-1) . A finite element basemat and superstructure model with smaller element sizes, i.e. , finer mesh, as compared to those used in the original basemat analysis (Ref. 2-5) wa s first ,,
- developed. The superstructure was included in the finite ~
element model i'n order to shnulate the stiffening ef fects of the major structural elements and these ef fects on the soil pressure distribution under the basemat. The soil surrounding the bottom of the basemat was represented by
! 3-directional soil springs.
Loadings which are consistent with those used in the origi-( nal basenet analysis were applied to the finite elenent model as point load, line load, or pressure load. Seismic loads were applied such that gross shear forces at the base-I mat level will be distributed in peqportion to the shear wall stiffnesses. These shear forces were applied along the respective shear walls at an appropriate height such that the resulting moment at the basemat level envelopes the
, gross moment obtained from the seismic soil-structure inter-l action analysis of the Reactor Building. Additional shear l forces resulting f rom the torsional moments were not consi-dered in the analysis as they would have negligible ef fect on vertical forces exerted on the basemat. Lateral soil pressure was not considered in the analysis since its ef fect is negligible for the basemat design.
Static analyses of the finite element ba s emat a nd su pe rs t ruc-ture were performed fci the loads and load conbinations as described in Secticas 2.1 and 2.2. The analysis res ults from the three directions of earthquake loadings were com-bined in accordance with "the comconent f a ctor me t ho d" (Ref. 3-2). The resulting moments, shears, and axial forces were used to determine the adequacy of the shear and flexural reinforcement at critical sections as well as the construc-tion joints of the basemat. In areas where the principal stress orientation significantly dif fers ' fran ~ the finite
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element grid orientation and/or the reinforcing grid systems, the sheGr twisting moments were accounted for using the
( CECAP canputer programs (Ref. 3-3).
4.0 Parametric Study In the original basemat analysis, only three brick elements were used through the thickness of the ba s ema t . The pur-pose of the parametric study wa s to detennine the number of brick elements to be used through the thickness of the ba sema t to provide sufficiently accurate element stresses to obtain appropriate moments.
An 160' x 160' x 14' concrete mat subjected to a point load of 4,000 kips at the cente r was analyzed using the finite element method. This configuration was selected A as it adequately represents the behavior of the Reactor Q1 Building basemat. Because of symme try , only one qua rter of the concrete mat wa s modelled in the analysis. Three finite element models with 3, 4, and 5 brick elements -
respectively through the thickness of the concrete mat were analyzed ( Figures 4-1 to 4-3). The nodal point. and element numbers of the basemat model are given in Figure 4-4. Figures 4-5 to 4-9 show the shear and normal stress distribution 'along the depth and width of the ba s ema t .
The analysis results show good agreenents between models with 4 and 5 brick elements through the thickness of the basemat. However, to provide better accuracy, 5 brick k elements through the thickness of the basemat were selected for tne analysis.
5.0 Finite Element Basemat and Supe rs tructure 11odel The finite element ba s ema t a n d su pe rs t ruct ur e no del in-cludes the 14-f t basema t, the dqrwell podestal, the cylindrical shell, and the interior and the exterior walls of the Reactor Building. Figures 5-1 and 5-2 show the isometric view of the basemat as well as the super-structure. The North-South and the East-West cross-sections of the model are show. in Figures 5-3 a nd 5-4, res pe ct ively. The top elevations for major structural components included in the model are also indicated in these figures.
The drywell pedestal and the basema t are nodelled by eight-node brick elenents. There are f ive laye rs of brick elements through the thickness of the ba sema t.
The shear walls including the cylindrical shell, the exterior as well as the interior walls are modelled by plate elements. As the brick element does not have the rotational degree of freedom , the pla te elements that simulate the walls are extended to the bottom of the ba sema t to ensure a reasonable stress distribution. The
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drywell shield wall above Elev. 83'-0" is represented by
( equivalent beam elements. Fictitious beam elements are also used to facilitate the application of line loads on structural walls. Slabs be tween walls are not modelled because the stress distribution due to structural stif fness has already been accounted for in the seismic soil-struc-ture interaction analysis of the Reactor Building . The model consists of 2,269 nodal points,1,366 brick elements, 7 50 pl a te el eme n ts ( Fig ur e s 5-5 to 5-2 3 ) a nd 149 beam elements . (Figure 5-24 ) . A plan view of the model at Elev. 4 0'-0 " is shown in Figure 5-2 5.
The foundation soil is modelled by attaching one vertical and two horizontal springs to each node at the bottom of the basemat. There are 303 vertical springs and 606 horizontal springs. The gross foundation spring constants
. are determined using the average final iterated soil properties obtained from the finite element seismic
- soil-structure interaction analysis of the major plant s truc tures. The gross spring constants are then distributed -
to each node in proportion to its tributary area.
Summaries of dead load, operating live load , and live load of the Reactor Building and the Radwaste Area are given in Tables 5-1 and 5-2. Ref e r to Figur es 5-2 6 and 5-2 7 for designation of walls and columns. Tables 5-3 a nd 5-4 show the horizontal North-South and East-West OBE
( seismic load for both the Reactor Building and the Radwaste Area. Additional vertical nodal loads due to North-South OBE earthquake are applied at the east and the west walls of the Reactor Building such that the resulting i moment at basemat level envelopes the gross moment obtained f rom the seismic soil-structure interaction analysis of the Reactor Building (Table 5-5).
I The maximum downward SRV discharge load per column is 755 kips. The maximum upward SRV discharge load per column is 813 kips. .
6.0 Analysis Performed and Results Static analysis of the finite element basemat and super-structure model was first pe rfo rmed usi ng BS AP for e igh t primary loads (Load Cases 1 to 8, Table 6-1) and five load combinations (Load Cases 9 to 13, Table 6-1). It should be noted that for Load Cases 9 to 13, the operating live load includes either upward or downward SRV discharge l o ad s . Also, the seismic loads obtoined from three direc-
- tions of earthquake were combined using the absolute sum me t hod . The finite element force / stress output from BSAP
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NW 7 85:.2 91510 is for each individual brick element which models the
( basemat. The resultant shear and moment due to the collective action of an individual stack of five brick elements through the thickness of the basemat were computed using the RESULT module of the BSAP conputer program. A detailed review of the resultant shear and moment outpu t indicated that Load Cases 12 and 13 are the gove rning load combinations. It is evident that the highest stress concentration occurs in the elements adjacent to the drywell pedestal.
To obtain more realistic analytical results, the seismic loads, Eo, for Load Cases 12 and 13 were recomputed u sing "tL canponent factor me thod" . This resulted in six additional load combinations, i.e. , Load Cases 14 to 19 ( Table 6-1 ) . The RESULT module of the BSAP canputer program was again used to compute the resultant shear and moment of each individual stack of five brick elements through the thickness of the basemat. Except for a f ew c elements adjacent to the dqrwell pedestal, all computed basemat tensile stresses are below the rupture modulus of concrete.
Figures 6-1 to 6-5 show the typical contour plots of the resultant transverse shear and moment for Load Case 1 (dead load ) . Also, the resultant transverse shear which
( is greater and M g, which thanare 100 kip-f t,than greater the 1,000 resultant kip-fmoments, t/f t, a nd M(he resultant twisting moment, Mr , which is greater than 500 kip-f t/f t. are given in Figures 6-6 to 6-9 res-pectively. The maximum resultant moment is 6,009 kip-f t/f t and the maximum resultant transverse shear is 631 kip /f t.
Stress concentration occurs in those elements located adjacent to the drywall pedestal.
Figure 6-10 shows the contour plot of the ef fective ver-tical soil pressure underneath the basemat. The average ef fective soil pressure is approximately 4.0 kip / f t2 which is approximately equal to the net ef fective pressure of the Reactor Building . The vertical deflection plots of the basemat under dead load are given in Figures 6-11 a nd 6-12. The deflection pattern of the ba sema t is
- similar to that of an upwardly concaved plate. Fig ure s 6-13 to 6-18 show the vertical deflection contours for Load Cases 14 to 19. These vertical deflection contours generally reflect the direction of the applied governing seismic load.
7.0 EVALUATION OF BASDtAT DESIGN The following itens we re evaluated to verify the adequacy of the basemat design.
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(1) Shear Reinforcement k The shear strength of the basema t is evaluated ba sed on the following conditions ( Fig ure 7-1 ):
- a. The critical section is selected at a distance of d/2 (where d is the ef fective depth of the basemat) fran the face of the drywell pedestal as allowed by ACI 318-71 for a two-way a ction of a slab. The nominal pennissiMe shear stress carried by concrete vc, is 4/f 6 where fh is the specified ninimum compressive strength of concrete (4,000 psi).
- b. One-way action is conse rva tively considered for jf the half circle of the cylindrical wall across the width of the basema t. The shear strength of the basemat is in accordance with ACI 318-71.
(2) Flexural Reinforcement The main reinforcement is reviewed for flexural and axial loads. The capacity reduction f actor is a f unction of the compressive force acting on each section. The minimum value used is 9 = 0.84. The required re inf o r ceme n t is compared with the reinforce-k ment provided .
(3) Construction Joints Shear reinforcement using No.10 rebar is provided in a 26" x 52" grid pattern throughout the Reactor Building basemat. For the Radwaste Area, shear rein forcement is not provided except along the column lines (Ref. 7-1). The horizontal cons truc tion joint is located at the mid-height of the ba s ema t .
The vertical construction joints for both the upper and the lower lif ts are shown in Figure 7-2.
The construction joints are prepared in accordance with Hope Creek Specification 10855-C-103(0) and in conformance with ACI 3 18-71.
The adequacy of the vertical construction joint is evalua ted at the location where the vertical cons-truction joint intersects the diagonal crack line a t mid-height. Using the shear friction approach, it is found that reinforcement perpendicular to the vertical construction joint plane is adequate to en-sure the proper transfer of vertical shear. Therefore, the ef fective depth of the basema t is utilized to resist shear. The vertical shear is then checked in
, [ accordance with the appropriate sections of ACI 318-17.
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{ The maximum horizontal component of the transverse shear occurs approximately at the midheight of the basemat where the horizontal construction joint is locat ed . Full transfer of horizontal shear is accomplished through shear ties and the concrete contact surface. The horizontal shear at the con-struction joint is checked in accordance with ACI l 318-71, Section 17.5 in which the basemat is treated as a conposite member with vertical shear stirrups (ties). Since the contact surfaces are clean and l intentionally roughened, the permssible horizontal shear stress (vh) is 80 to 350 psi depending on the amount of shear ties provided.
I (4) Thermal Leads The ef fects of thermal gradients through the basement thickness result in the requirement of 0.8 in2/f t of reinforcement (Re f. 7-2 ) . For load combinations in which thermal load la included, all the load f actors are reduced by one quarter (Ref. 2-2). There fore ,
the thermal load is adequately accounted for by the existing design.
(5) Twisting Moments
( The twisting soments were accounted for using the CECAP computer program (Ref. 3-3 ) .
For items 1 to 3, the evaluation was performed for the following four general areas of the basemat.
- a. Around the drywell pedestal,
- b. Between drywell pedestal and cylindrical wall,
- c. Around cylindrical wall, and
- d. In the radwaste area.
As can be seen from the summa ry of evaluation results presented in Table 7-1, the design of the basema t neets all applicable code req ui reme n ts.
For item 5, typical critical brick element stacks at or near the corner of the basema t and at the base of the drywell pedestal were selected for evaluation. As indi-ca ted in Table 7-2, all the camputed maximum reinforcement and concrete stresses are less than the allowables.
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L APPENDIX A R EFERENCES G
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REFERENCES )i k
1-1 Final Report, " Hope Creek Generating Station, Indepen-dent Design Verification Program," prepared by Sargent and Lundy Engineers for Public Service Electric and Gas Company, August 30, 1985. )
l' 2 OR 7, Independent Design Verification Program Observation Report No. 7, Rev. O , M ay 18, 1985.
1-3 OR 4 0, Independent Design Verification Program Observation Report No. 40, Rev. O, May 28, 1985. l 2-1 Specification 10855-C-103(O), " Technical Specification for Fo rmi ng , Pla ci ng , Finishing and Curing of Concrete for the Hope Creek Generating Station," Rev.13. I i
2-2 Specification 10855-D2.1, " General Civil-St ructural Design Criteria for the Hope Creek Generating Station, -
Public Service Electric and Gas Company, Newark, New Je rsey , " Rev . ,8.
2-3 " Seismic Structural Analysis, Final Report, Hope Creek Generating Station", prepared by Impell Corporation for Public Service Electric and Gas Company, Impell Report No. SED-76-017, Rev. 5 , De c embe r 19 83.
k 2-4 5 Hope Creek Generating Ststion, Plant Unique Analysis Report," prepared by NUILCH Engineers, Inc. San Jose ,
California for Public Service Electric and Gas Company, PBC-01-3 0 0-1, Rev. O , Ja nua ry 19 84.
4 2-5 Design Summary " Foundations for Reactor and Auxiliary Building for the Hope Creek Generating Station, Public Service Electric and Gas Company , Newa rk , New Jersey,"
Rev. B, March 28, 1976.
l 2-6 " Impedance Approach Seismic Soil-Structure Interaction Analysis of Category I Structures," prepared by Bechtel Power Co'poration, r San Francisco, California for Hope
. Creek Generating Station, Public Service Electric and Ga s C ompay , Rev . O , J uly 19 84 .
3-1 CE800, BSAP, Bechtel Structural Analysis Program, Release E16-50, August 17, 1984.
3 -2 Design Criteria No. C-2.44, " Seismic Analysis of Strctures and Equipment for Nuclear Power Plants," Rev. O, Augus t 1980.
4 3-3 CE987, CECAP, Concrete Element Cracking Analysis Program, Program Release A7-7, March 30, 1984.
(
F90/3 A-1
N07 7 8502 91510 REFERENCES (Cont'd) 7-1 Hope Creek Generating Station, Civil Drawing Numbers C-0481-1, Rev. 6, C-0 4 8 2-1, R ev . 3 , a nd C-0 4 8 3-1, R ev . 8 ,
" Foundation Mat Reinforcement Detail, Sheets 1 to 3".
7-2 Hope Creek Generating Station, Civil Calculation Number C621-16 ( Q ) , Rev. 2, " Reactor Building Foundation Rebar Requirements".
t e
e
~
l l
l 1
(
F90/3 A-2 l
r
APPENDIX B TABLES e
Table 5-1
SUMMARY
OF DEAD LOAD (DL), OPERATING LIVE LOAD (Lo)
( AND LIVE LOAD (L) - HEACTOR BUILDING DL i Lo l L Structure Accumula ted Accumula ted Accumula ted Component 0 El 54' @ El 54 ' @ El 54' i ]
l Cylind rical Wall 95,855.7 4,964.1 23,296.4 1
South Ex-terior uall 13,516.2 215.1 933.2 North Ex-terior Hall 9,713.1 215.1 922.9 -
l West Exte-rior Hall 12,786.2 177.5 762.3 East Wall 9,511.5 177.5 762.3 l
( -
Interior Wall 4,716.1 299.0 1,265.0 Drywell Shield 96,683.4 1,928.7 7,476.5 l
Toru s 9,763.0 29.4 1 17 .5 LOCA Flooo on Shield (Post 9,550.0 - -
Accident Flooding) __
Post Acci-dent Flood- 9,087.0 - -
i ng Toru s Note: All Units are in kips
(
B-1 F82/8
N07 7 25 G 2 91510 Table 5-2
(. Summary of bead Load (DL) , Operating Live Load ( Lo)
And Live Load (L) - Radwaste Area DL L Lo Structure Accumulated Accumulated Accumulated' Component 9 El 54' 9 El 54' 9 El 54' Wall A 4994.0 824.1 206.0 Wall B 14288.0 1756.8 495.4 Wall C -
Wall D - 15589.0 3071.1 832.3 Wall E 2052.0 1215.1 425.3 Wall F 5740.0 551.5 164.9 Column 1 1084.0 663.0 165.8 Column 2 1168.0 718.4 179.6 Column 3 2620.0 908.2 227.1
( 3037.0 643.5 160.9 Column 4 Column 5 2812.0 850.3 212.6 Column 6 2785.0 993.4 248.4 Column 7 1294.0 720.8 180.2 Column 8 1109.0 664.9 166.2 Column 9 1188.0 362.7 90.7 Column 10 1155.0 482.5 120.6
]
ll Note: All Units are in kips.
1 F82/8 B-2
Table 5-3
( Horizontal North-South OBE Seismic Nodal.Ioads RADESE AREA REACIOR BUIIDING EAST WALL WEST WhLL EAST WALL WEST WALL CYLINDER WALL (El. 147') (El. 147') (El. 132') (El. 132') (El. 212')
Nodal Ioad Nodal Ioad !bdal Inad Nodal Ioad Pbdal Ioad Point Kips Point Kips Point Kips Point Kips POnt Kips 8293 292.10 8249 202.33 8238 259.58 8212 327.78 8322 501.89
- 8294 584.20 8250 404.67 8239 522.66 8213 792.50 8323 501.89
, 8295 584.20 8251 445.36 8240 560.08 8214 1163.31 8324 501.89 8296 584.02 8252 404.67 8241 526.16 8215 1005.84 8325 501.89 _
8297 584.20 8253 393.97. 8242 492.25 8216 594.51 8326 501.89 8298 617.17 8254 427.47 8243 526.16 8217 594.51 8327 501.89
. 8299 650.13 8255 450.26 8244 492.25 8218 594.51 8328 ' 501.89 8300 650.13 8256 450.26 8245 526.16 8219 615.40 8329 501.89 i (
8301 643.03 8257 445.55 8246 560.08 8220 636.28 8330 501.89 l
l 8302 636.42 8258 440.84 8247 540.20 8221 899.80 8331 501.89 8303 318.29 8259 220.42 8248 277.11 8222 1163.31 8332 501.89 8223 941.57 8333 501.89 l
8224 359.92 8334 501.89 8335 501.89 8336 501.89 l
8337 501.89
- - - - - - - - 8338 501.89 8339 501.89
(
F82/8 B-3
, 3,.
_ ... ,,.. ,,.. F, I
244 345 344 347 42M 9210 8178 Sitt SISO See 348 842
, 343 k
sin .2,. ,,r. ,i.. ..
SH Fee FSI F02 J2M 3210 3170 3189 3150 l
l
( ... ... ..
2M 210 179 189 150 SECTION 0-0 t
i FIGURE 5-19 l
i
, C-34 i
t
- - . - ~ .__ _ _ . _ _ _ _ _ _ _
Table 5-4 Horizontal East-itst OBE Seismic Pbdal Loads RADWMTTE AREA REACIOR BUIIDING IORIH WML SOUni)RLL NOR1H WALL SOUIH WEL CYLINDER WM L (El. 120') (El. 120') (El. 120') (El. 120') (El. 130')
Nodal Ioad Nodal Ioad tedal Ioad Nodal Ioad tbdal Ioad Point Kips Point Kips Point Kips Point Kips Pont Kips 8293 255.32 8303 468.26 8238 320.15 8248 586.39 8130 622.44 8282 510.65 8292 936.53 8196 503.32 8195 921.88 8131 622.44 8271 458.82 8281 841.48 8197 366.30 8194 671.10 8132 622.44 8260 407.01 8270 746.46 8198 366.30 8193 670.86 8133 622.44 ,.
8249 203.50 8259 373.22 8199 366.30 8192 670.86 8134 622.44 8200 520.31 8191 953.00 8142 622.44 8201 674.17 8190 1234.82 8143 622.44 8202 674.17 8189 1234.82 8144 622.44
(
8203 674.17 8188 1234.82 8145 622.44 8204 538.42 8187 986.16 8146 622.44 8205 402.65 8186 737.51 8147 622.44 8206 402.65 8185 737.51 8148 622.44
_- - - - 8207 402.65 8184 737.51 8149 622.44
- - - - 8208 402.65 8183 737.51 8150 622.44
- - - - 8209 374.49 8182 685.92 8158 622.44 8225 651.08 8237 1192.52 8159 622.44
- - - - 8212 477.92 8224 875.37 8160 622.44
- - - - - - - - 8161 622.44
(
F82/8 B-4
^ -
Table 5-5 k Added Vertical Nodal Loads Due To Horizontal North-South OBE Earthquake - Reactor Building EAST WALL WEST WALL Nodal Force Nodal Force Point Kips Point Kips 8238 -602.9 8212' - 783.2 8239 -1080.0 8213 -1892.8 8240 -853.3 8214 -1860.9 8241 .-549.0 8215 -996.9 8242 -244.7 8216 -397.6 _.
8243 _9 . 0 8217 -210.4 8244 227.9 8218 -23.3 8245 531.0 8219 176.6 8246 805.7 8220 389.8
[
8247 1107.0 8221 980.0 8248 667.3 8222 1821.3 8223 1943.6 8224 853.8 i -
F82/8 B-5
TABLE 6-1 LOADS AND LOAD COMBINATIONS
(
l 'l l Load Case Load Description Note l
- 1. Dead Load f ran Buildi ng IJalls, Equipmme nt i and Component on Elev . 5 4 '-0 " , t o ru s, -
and Buoyancy (DL)
- 2. Operating Live Load f rom Building llalls, and Elev. 54'-0" (Lo) -
l
- 3. Live Ioad f ran Building tialls and Elev.
54'-0" (L) -
l
- 4. Post Accident Contairraent Flooding Includ-l ing Torus and Drywell Shield LJall (F) i -
l
- 5. Horizontal North-South Seismic (OBE) Load j from Duilding llalls, Ba sema t, a nd Toru s -
(Eo,NS) ,
( 1
- 6. Horizontal East-lJest Seismic (OBE) Load from Building Italls, Ba sema t, a nd Toru s -
(Co , Ell) l
- 7. Vertical Seismic (OBE) Load f ran Buildi ng l lialls , Basemat, a nd Toru s ( Eo ,y ) ,
l
- 8. SRV Dicharge load ( S RV ) -
l l
- 9. 1. 4DL + 1.7L + F -
- 10. DL + Lo + Eo + F 1,3
- 11. DL + Lo + Eo + F 2,3 1
- 12. 1.4DL + 1.7 Lo + 1.9 Eo 1,3
~
(
i l
F82/5 B-6
TABLE 6-1 LOADS AND LOAD COf1BINATIONS (Con t ' d )
k-l l l
Load Case Load Description Note
- 13. 1.4DL + 1.7 Lo + 1.9 Eo 2,3 l
- 14. 1.4DL + 1.7 Lo + 1.9 Eo 1,4
- 15. 1.4DL + 1.7 Lo + 1.9 Eo 1,5
- 16. 1.4DL + 1.7 Lo + 1.9 Eo 1,6
- 17. 1.4DL + 1.7 Lo + 1.9 Eo 2,4 _
- 18. 1.4DL + 1.7 'Lo + 1.9 Eo 2,5
- 19. 1.4DL + 1.7 Lo + 1.9 Eo 2,6
(
NOTES: 1. Lo includes downward SRV load.
- 2. Lo includes upward SRV load.
- 3. Loads from three directions of earthquake are combined using the absolu te sun me tind.
- 4. Loads from three directions of earthquake are ccrabined using the ccznponent factor method.
1.0 Eo,NS + .4 Eo,Eti + .4 Eo ,y
- 5. Loads from three directions of earthquake are combined using the component factor me thod.
.4Eo,NS + 1. 0 Eo, Eli + .4 Eo,y I
- 6. Loads from three directions of earthquake are combined using the ccrnponent f actor me thod.
.4Eo,n3 + .4E o ,Eli + 1. 0 Eo,y
(
F82/5 B-7
m " r.
TALLE 7-1 SUtt1ARY OF REM.*IOR HUILDING ImSEMAT DESIGJ INALUATION O e Hain Reint. In /Ftl Shear Heinforement Requirtti Transv. Shear florizontal Shear l ARFA Pro- tiax . lAllw Hax . lA llm* lReint . l Heint, , RB1AIES Bending vided Stress l Stress Stressj Stress l 'd Prwidel l PSI PSI l PSI PSI Raj/Ft lIn 2 In2/Ft 2 '
l l l
- Arcurnt Dryell 7.34 7.38 240 253 240 350 .12 . 135 -
Pedestal l l 1 I I uetween ury-wil 1%)es- l
, tal & Cylin- 5.52 7.38 223 253 Not Critical -
l . drical uall l i Alorg 4.55(1) l 5.54(1)l 103 115 tbt Critical Cylindrical 0.86( ) 3.69( ) - -
Wall j l j 101 ,
Horizontal shear allowable 1
Rad e sta 1) 3.69(2) 92 126 73 to Varies l Varles stresses are Iroratal fron l Area 3.60(2) 1.85( 5.54( ) 229 l AC[ values accordirri to j l l inforement prwided. l fUILS:
See ACI 318-71 (17.5)
(1) Reinforcement in East-West Direction. -
(2) Reinforetrnent in North-South Direction.
1 i
199/9 .
s L
i NW 7' esc 291510 TAILE 7-2 Maximm Reinforcing and Cbncrete Stresses frm CECAP Analysis Brick l )
Elenent! Maximm ReinforosnentllAllmable Reinforcement Maximm Concre Stack Stress (ksi) ( Stress (ksi) l Stress (ksi) l Concrete Neber I Stress (ksi)' !
1 1 l l
21 '
51.6 l 59.7 1 2.59 l 4.30 l l \ l 23 ! 55.9 57.6 2.22 l 4.15 l l l l 28 l 50.3 1 59.7 i 2.47 l 4.30 l '
l 29 ! 55.0 61.0 1 1.68 l 4.39 1 1 ,
\ .
270 1 25.3 l 60.3 '
O.82 1 4.34 1 s
NOPIT.S 1. Allowable reinforcenent stress is capacity reduction factor
(#) times in-situ reinforcenent strergth.
- 2. Allowable concrete stress is capacity reduction factor
(#) times in-situ concrete strergth.
4 1
1 F89/9 B-9
.- . . . , ~ , , . - , , , . . ,_._,,,+y., . - ,.-,_-.. - ,_ ,. , . - _ . - , , _ , .
.m *=
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APPENDIX C FIG URES G
F90-3 i
( ,-
2 N=
I 5a Es 1I
=
630'O' s
l _
265'6' - _
364'6' _
ts 11 ? ADMINISTRATION TURBINE
~
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- FACILITY BUILDING sx ~
1! .
]I i
-n s .
^
- j h cA$I ARE BUI' I I I bh Et a, - -
I
%j 19 2 ' 6 " 19 2 ' 6 "
g = _
16 5 ' 0 " .
550'o-2-
s1 -
GENERAL PLANT LAYOUT
.h ly FIGURE 2-1
- I I5 l <
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f ,g1eg i!$
@ C-1 h
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l UPPER BASE MAT R OCKS
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4 1 sE :
{2 $ 51 66 55 57 32 37 11 22 4
) I' 64 62 65 49 40 24 28 17 6 E g 50 67 56 63 35 36 13 20 3
- .~
45 54 46 27 23 9 4 31 10 61 48 60 30 39 19 M 18 g
g g 43 53 44 34 26 8 21 7 3 g 59 47 58 29 38 16 12 15 I 42 52 41 33 25 2 5 1 _.
sx .
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ifn-y R_
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- 96'3'_
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2 f h BRICK ELEMENT NO.
II y
$2 h 1} 31 3 35 36
}{ 80.0, 52 U O tj @ @ @ @
5 25 26 27 28 29 30 h 5 7.12 '
it sr '
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11 I 3 24 43 37.I'
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*b 0.0' 7.14 ' 20.195' I *1 II f
- E m
( ee = PLAN VIEW OF BASEMAT MODEL FOR PARAMETRIC STUDY
; FIGURE 4-4 3E@a c-6 h
m l 053:C103020]HCGS2.DGN PF=l6
e 5 LAYER MODEL g E 4 LAYER MODEL A 3 LAYER MODEL E E* 3} E x gk 11 . k is- - 3C sr . 13 C #
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&j i TRANSVERSE SHEAR STRESS DISTRIBUTION ALONG BASEMAT DEPTH, .
]f[ AT CENTROID OF BRICK ELEMENT STACK NO.I 45 FIGURE 4-5 g
a .
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_ _ , _ _.053:[103020]HCGS3.DGN _ _ PF=0i _ _ . __ _ . _ _ _ _ _ __ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _
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'R $2 .
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=, ,s s s' k
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&{*
3
- y _3 AT CENTROID OF BRICK ELEMENT STACK NO.2
( FIGURE 4-S o{g 8IO '
$ C-8 m
053:[103020]HCGS3.DGN PF=:1
e 5 LAYER MODEL E 4 LAYER MODEL
} A 3 LAYER MODEL 52 Es III '
d g i I 15 - -
~
it Ek 1 - Il I 11
*I W 10 -
k{ G li 5 II E 4 gr s g3 5- - \ I g
~
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El gb o 0' s'N 20' 40' 60' 80'
- =
gI DISTANCE IN FT (FROM LOAD APPLICATION) NOTE: DIFFERENCES BETWEEN DATA POINTS ARE TOO SMALL TO DISTINGUISH ON THIS PLOT I l 1f< gf TRANSVERSE SHEAR STRESS DISTRIBUTION ALONG BASEMAT t c, FIGURE 4-7 b w C-9 OS3:[iO3020]HCGS3.DGN PF::1
l ( e 5 LAYER MODEL 5 4 LAYER MODEL A 3 LAYER MODEL g.E - 1I . 1,x h Il
~
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'a '
s f a'
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"k -20 -10 0 10 20 30 l} NORMAL STRESS, KIP /FT2
*I 1]
E
]! NORMAL STRESS DISTRIBUTION ALONG BASEMAT DEPTH,
({[$ jo AT CENTROID OF BRICK ELEMENT STACK NO.I FIGURE 4-8 iQg . C-10 W -. . . - - _ - . - - . - . _ _ _ _ - - - - _ _ .
e 5 LAYER WODEL E 4 LAYER MODEL y y A 3 LAYER WODEL J 5j \ E2 3I 1as b h I1
~
30--T At ~ g ss - II b I l 11 I i 20- - i fg l 1 < 0 EI l b ti w
*g 10 -- \
g i i R*1 20' g] o , x_____ __
=
" is O' 40' 60' 80'
- DISTANCE IN FT (FROM LOAD APPLICATION)
=
II l I g{o I {f, y NORMAL STRESS DISTRIBUTION ALONG BASEMAT FIGURE 4-9
$$h b ?" _ _ _ _ -_ _ _
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s
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l' -dW'C1 m + _ _ _ _ - s
/X 9
Notes g - global coordinate system AN ISOMETRIC VIEW OF THE BASEMAT 1
- FIGURE 5-1
, i k
l C-12 l
e 4 m ham- =m
@6M E @=# # me 4 +Ge@ee9 se &
]
( newwweun M\'MM\W\X\\y\
~
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\V
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=
y y M /^'//5 5 \Vd'\ 44 ( M/MbVTTviv" Mt/V/ viv
= s na-
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=-
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d Y _ w -- w _ Sr . es
\ \ \ '\ h g"g '~ (-~
AN ISOMETRIC VIEW OF THE SUPERSTRUCTURE FIGURE 5-2 I c-13
LEGEND: n BEAM ELEMENT
. PLATE ELEMENT
_( 3l =
= BRICK ELEMENT 11 5a ' $2 .
ELEVATION (FT) is ne s '
}I I
g . a - AF . sr . : 13 0 11 : 20 - (! Og 5 El J1 m- i IrI :
*g :
83 - gf 7T - a ik
-I 54 - --
I < I ' 2 40 - ( j5 - NOTE: NOT TO SCALE o1o N-S SECTION OF BASEMAT AND
$5b SUPERSTRUCTURE MODEL k FIGURE 5-3
$ C-14 m rn a
l ELEVATION FT) LEGEND: l CZC223 BEAM ELEMENT
- PLATE ELEMENT 2 C BRICK ELEMENT 1
- 5a
$2 I .
E 2 k
. 1 1
} IAT - _.
- at -
sr . l IJ 132 - 11 (1 km ils - . El : El i gg 80 2 - B3
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{f gl as - TT - El gh 81
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E . Ifg, do - I S- NOTE: NOT TO SCALE E g E-W SECTION OF BASEMAT AND I O SUPERSTRUCTURE MODEL
$ FIGURE 5-4
- u. C-15
_ _ _ i m w a m m e w u w m_____ m a a - _ _ . -- - . - _ - _ -
! m < O Z
< 'l _ i-* F _.
sh 12
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- 1j
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i ATTACHMENT 3
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SG CHIEF PROJECT ENGINE
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a mno -- Public Service Electric and Gas Company P. O. Box A Hancocks Bridge, New Jersey 0$038' #
"C'/ 2 G E9C Hope Creek Generating Station ry;-
V E'X g SYST-Y NOV 2 51985 SEE E-d' .Ri .K4
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File 403.2 (SE.85.11.20-10) To the Chief Project Engineer - Hope Creek REVIEW OF TRANSMISSIBILITY FOR ALL NUCLEAR SAFETY RELATED LOCAL PANEL MOUNTED DEVICES
REFERENCES:
- 1. OBSERVATION / RESOLUTION / COMPLETION REPORT NO. 145.
INDEPENDENT DESIGN VERIFICATION PROGRAM, IDVP
- 2. NRC IDVP FOLLOW-UP AUDIT EXIT _
MMETING MINUTES (BLP-18,061) HOPE CREEK GENERATING STATION During the period of November 12 through 15, 1985, Mr. G. Luh and Mr. L. Tao of Site Engineering, took s trip to General Electric office at San Jose, California to revicw the seismic transmissibility of GE supplied safety-related local panel mounted devices at Hope Creek Generating Station. This report documents the close out on the subject action item as identified in NRC IDVP Follow-Up Audit Exit Meeting (See Discussion Item No. 8 of Reference 2). All GE Seismic Qualification documents related to the subject were reviewed with the following comments:
- 1. SORT Report: " Seismic Qualification Reevaluation, Class lE Equipment," DRF A00-01397, Volume 2, Rev. 2, local panels.
o Hope Creek Unit I local panels and associated Class lE devices were seismically qualified through type testing and similarity analysis to other panels (30", 48", and 72" H22 panels) and instruments. o Rack-mounted instruments were qualified by testing with the instruments mounted on the panel and the Test Response Spectra (TRS) envelops the Required Response Spectra (RRS).
A Chief Project Engineer - 2 NOV 2 51985 Hope Creek o In the case where the instruments were not mounted on the test panels, qualification was based on the criteria that instrument TRS envelops the theoretical RRS at the instrument mounting location considering the panel transmissibility. o Devices which were removed from racks and then mounted locally were qualified by comparing the device TRS with the RRS which envelops all locations where the devices are mounted locally. o All devices located on local racks and were not being qualified by other programs (i.e., Phase III Program), ~ were shown as qualified to the SORT program criteria. o Page 9 of S0kT Report needs to be revised to reference appropriate figures. Table A of SORT Report needs to be revised for typo and to indicate the correct qualification method for certain instruments mounted on panels H21-P001 and H21-P030 through H21-P033. o HCGS FSAR Tables 3.10-3 and 3.10-4 need to be revised to categorize all Class lE devices into three groups namely shipped loose devices, local panel devices, and control panel devices.
- 2. EO Phase III Report, Book No. C-59 " Environmental Qualification Report for Pressure Transmitters", GE Drawing 188C7360, Rosemount Model No. 1153 Series B, Rev.
2. o Rosemount Model No. 1153B Series pressure transmitters were primarily qualified by type test following the guidelines of IEEE Standards 323-1974 and 344-1975. Supplementary Analysis were also used to show the qualification of untested nodels based on similarity to those tested. Tests included seismic aging and seismic abnormal testing. o RRS for 30" panel (Ref. H21-P014) needs to be added to Book No. C-59. o Page 25 of Production Evaluation Section (Book C-59) needs revision in the Seismic Qualification paragraph to reference correct figures.
f Chief Project Engineer - 3 NOV 2 51985 Hope Creek
' 3. EO Phase III Report, Book No. C-13, " Environmental Qualification Report for Rosemount Pressure Transmitter,"
Rev. 2. o The transmitters in this group are predominantly local panel / rack mounted Rosemount 1151 transmitters which were supplied with an optional mounting bracket for either panel or 2" pipe mount.
- g. _
o The transmitters were shown to be family members. A
-similarity analysis showed that there are no design
' differences due to service application, and the variations in stresses and materials are insignificant ~..
and do not affect safety related functions. o Both type test and similarity analysis were performed to show that the devices were seismically qualified for safety related functions. o Book No. C-13 needs clarification on TRS values used in qualifying the transmitter by test, and the acceleration level of 20 g resulting from the analysis.
- 4. EO Phase III Report, Book No. C-25, " Environmental
-Qualification Report for GE voltage preamplifier."
o Listed under MPL C51-K002A-H, the GE voltage preamplifier was qualified seismically by teFt. The device, while mounted in a H22-P030 Hoffman enclosure, was qualified to OBE and SSE levels which exceeded our requirements. o Table A of SORT Report shows that C51-K002A-H were mounted on H21-P030-P033 and were qualified as Rack Mounted following SORT program criteria. However, Book No. C-55 indicated that C51-K002A-H for P030-P033 panels were qualified under EQ Phase III program. This has to be clarified. In summary, the methodology, testing values, and analysis results used in qualifying Class lE equipment were reviewed and found to be satisfactory and acceptable. All Safety Related local panel mounted devices were seismically qualified for their respective mounting locations, and the
Chief Project Engineer - 4 Hope Creek NOV 2 51985 transmissibility had been addressed adequately. This report confirms the statement made in the response for OR No. 145.
) j e d . I, m u t v C. W. Churchman Site Engineering Manager -
Hope Creek h GL:thj C A. S. Kao SE File STAIRS i i 4 j
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ATTACHMENT 4 l i i December 18, 1985 File 403.2 (SE 8 5.12.17 ) To the Chief Project Engineer - Hope Creek COMPLETION REPORT ON FLANGE BOLT REVIEW INDEPENDENT DESIGN VERIFICATION PROGRAM
REFERENCES:
- 1. OBSERVATION / RESOLUTION / COMPLETION REPORT NO. 144, IDVP
- 2. NRC IDVP FOLLOW-UP AUDIT EXIT MEETING
, MINUTES (BLP-18,061)
HOPE CREEK GENERATING STATION Per Reference 1 above, PSE&G and BPC agreed to review the , flange bolt design for non-NSSS active pumps. This report documents the completion of the review and forms a closecut on the subject action item as identified in the NRC IDVP Follow-up Audit Exit Meeting (see Discussion Item No. 8 of Reference 2). PSE&G has reviewed calculations (678-88(0) and 678-95(0)) provided by Bechtel. All flange bolts for non-NSSS active pumps as listed in FSAR Table 3.9-18 were reviewed for the design adequacy. Pumps with revised nozzle loads (e.g., SACS pumps) were re-evaluated and found that bolt stresses calculated using the actual piping reaction are less than the allowable values. For other pumps, evaluation was done based on the faulted condition loads. In this case, nozzle loads used in the analysis are the maximum allowable faulted loads. Results show that the flange bolt stresses are within the allowable limits. Therefore, it is concluded that all the HCGS non-NSSS active pump nozzle flange bolts are adequately designed and meet the FSAR and appropriate technical specification requirements. b- b G. G. Luh
/[h-k GL/lb Senior Staf f Engineer PE15/34 C A. S. Kao M. C. Reeser SE File STAIRS
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