ML20045G804
| ML20045G804 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 05/07/1993 |
| From: | Hookway R EQE, INC. |
| To: | |
| Shared Package | |
| ML20045G695 | List: |
| References | |
| 42094-P-003, 42094-P-3, NUDOCS 9307160010 | |
| Download: ML20045G804 (10) | |
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ENGINEERING ANALYSIS GUIDELINES CON SU LTA N TS FOR CONNECTICUT YANKEE ATOMIC POWER COMPANY SEISMIC UPGRADE PROGRAM FOR SERVICE WATER SYSTEM, FEEDWATER SYSTEM AND MAIN STEAM SYSTEM OUTSIDE CONTAINMENT Project Performed For:
Northeast Utilities Service Company PROCEDURE NO.:
42094-P-003 REVISION NO.-
1 DATE:
May 7,1993 1
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9307160010 930709 PDR ADOCK 05000213
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PDR te, EQE ENGINEERING CONSULTANTS A Division of EQE hiternational
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42094-P-003 Revision 1 May 7,1993 Page 2 of 10 TABLE OF REV;SIONS Revision -
No.
Descriotion of Revision Date Accroved 0
Original issue April 2,1993 1
Revised paragraphs 1.0 and May 7,1993 2.0. Also added paragraphs 8 & 9 to decribe Project inter-face and QA requirements. A vertical bar in the right margin identifies all revised sections.
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42094-P-003 Revision 1 May 7,1993 Page 3 of 10 I
TABLE OF CONTENTS I
Pace TABLE OF REVISIONS......
2 1.
PURPOSE..............................................................
4 2.
SCOPE,....................................................................
4 2.1 Service Water System Piping..........................
4 2.2 Main Steam System Piping...
5
- 2. 3 Fe e d w a t e r Pi ping.......................................
6 3.
MODELING/ TECHNICAL CONSIDER ATIONS......................
6 3.1 Single Acting R e straints.....................................
6
- 3. 2 R o d H a n g e r s.....................................................
7
]
1 4.
LOAD COMBINATIONS AND ACCEPTANCE CRITERIA.......
7 4.1 Piping 7
- 4. 2 Pipe Sup port.....................
8 5.
SEISMIC MA RGIN AN ALYSIS......................................
9 6.
REFERENCES........
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42094-P-003 Revision 1 May 7,1993 Page 4 of 10 I
1.
PURPOSE This procedure provides an outline of the criteria and methods to be employed for the seismic evaluation of the Service Water System, Feed-water System, and Main Steam System piping at Connecticut Yankee.
'I The methods included in this document are intended to supplement those presented in Reference 1 which is applicable to all safety related piping at the plant. That is, the guidelines of Reference 1 are to be followed except as modified herein.
2.
SCOPE 2.1 Service Water System Piping The piping included in the scope of this effort includes the appropriate supply and discharge piping as described on the following drawings:
20231-SH 105-AG 20231-SH-105-AH 20231-SH-105-AJ I
20231-SH-105-AK 20231-SH 105-AL 20231-SH-105-AM I
20231-SH-105-AV 20231-SH-105-AW 20231-SH 105 AX 20231-SH-105-AY l
20231-SH-105-AZ 20231-SH-105-BA 20231-SH-105-BC 20231-SH-105-SD 20231-SH 105-BH 20231-SH-105-BJ 20231-SH 105-BK 20231-SH-105-BM 20231-SH-105-BN I
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42094-P-003 Revision 1 May 7,1993 Page 5 of 10 All of this piping is in the Primary Auxiliary Building and a large portion is supported from the roof at Elev. 53'-8".
This piping consists of 16" & 18" header piping at Elevation 50'-0"in the Primary Auxiliary Building with numerous branch lines varying in size between 6" to 16" diameter. The branch connections consist primarily of unreinforced stub-in connections. Support for this piping consists of short rod hangers from the ceiling, lateral supports on the headers and stanchion supports for piping near the floor elevations.
2.2 Main Steam System Piping The piping included in the scope of this effort includes the appropriate piping from the Reactor Containment penetration to the stop valves at Elev. 66 W This piping is shown on drawings:
20231-SH-101 A 20231-SH-1018 20231-SH-101 C 20231-SH-101 D 20231-SH-101 E 20231-SH-101 F The 24" diameter Main Steam lines run from the containment penetration to I
the 35" diameter manifold header at Elevation 51' 6" in the Turbine building.
Branch connections of 12",18", 24" and 30" diameter on the manifold made with reinforcing pads on stub-in connections lead to the stop valves at Elevation 66'-6". This piping is supported primarily on spring hangers and sliding supports.
Only the short portions of piping (= 21') between the Reactor Containment and the Containment isolation Valves is safety related. Pressure boundary integrity must be assured for the safety related piping. The remainder of the piping is non-safety related and only failure and falling is of concern.
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7 42094-P-003 Revision 1 May 7,1993 Page 6 of 10 2.3 Feedwater Piping The piping included in the scope of this effort includes the appropriate piping from the Reactor Containment penetration to the Valve in the Turbine Build-I ing which is required to ensure Auxiliary Feedwater Flow. The piping down-stream of the this Valve is considered non-safety related. This piping is shown on drawings:
20231-SH-102A 20231-SH-102B 20231-SH-102C 20231-SH-102D 20231-SH-102E 20231-SH-102F 20231-SH-102G 20231-SH-102H 20231-SH-102J l
20231-SH-102K I
20231-SH-102L 20231-SH-102M The 12" diameter Feedwater line numbers 7,8. 9, and 10 run from the containment penetration to an 18" header pipe at Elevation 47'-13/8" in the Turbine Building. Branch connections of 12" and 18" diameter on the manifold made with reinforcing pads on stub-in connections provide a by-pass to the 12" Main Feedwater line and lead to the Feedwater Heaters.
Support for this piping is provided primarily by spring hangers and sliding supports in addition to an intermediate pipe anchor between the Reactor Containment and the Turbine Building.
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MODELING/ TECHNICAL CONSIDERATIONS 3.1 Single Acting Restraints As substantiated by past experience, piping system uplift is considered not I
to lead to credible failure modes for piping. Design margins in support systems are considered sufficient to encompass the potential support load increases due to the impact which could result from the landing of uplifted piping.
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O 42094-P-003 Revision 1 May 7,1993 Page 7 of 10 3.2 Rod Hangers For rod hangers, vertical uplift due to seismic response may be neglected, impact loads, as described above, need not be explicitly addressed.
Supports for which piping motions are limited by the physical design (e.g.
short rod hangers) may be modeled as springs with the spring rate chosen such that the resultant displacement is limited to the maximum possible displacement allowed by the design. This approach may also be used for other intervening structures such as floor and wall penetrations.
I 4.0 LOAD COMBINATIONS AND ACCEPTANCE CRITERIA i
4.1 Piping The structuralintegrity of the safety related piping in the systems as described in Section 2.0 will be assured by satisfaction of the following:
1-Pr + DW + ML < 1.0 Sh 11-Th < Sa Ill-Pr + DW + K (SME(l)2 + SME(SAM)21/2 < 3.0 S 3
h Where:
Pr Pressure Stress
=
DW Stress due to dead weight
=
ML Stress due to mechanical loads (as defined in the
=
design specification)
Th = Thermal expansion stress (with appropriate SIF)
SME(l) = SME inertia stress SME(SAM)
SME seismic anchor motion stress
=
S Yield stress
=
y i
Sh Allowable stress
=
SA f(1.25 Sc + 0.25 S )
=
h K
0.8 (ductility reduction factor) (See Reference 4)
=
g lim
y 42094-P-003 Revision 1 May 7,1993 Page 8 of 10 Piping system analysis calculations for determination of seismic response will I
consider 5% damping for floor response spectra as described in Section 5.0.
The analyses may consider the realistic effects of nonlinear behavior as appropriate including proximity / impact with other systems, interference and small clearances to stiff structures, geometric restoring forces, and wall penetration sealants.
4.2 Pipe Support I
Piping supports will be accepted based on stress allowables or test data as follows:
Acceptable flexural and tensile stresses are the lesser of 0.7 Su and 1,2, S.
y Acceptable shear stresses are the lesser of 0.42 Su and 0.72 Sy.
Acceptable bolt stresses are the lesser of 0.75 Su and the minimum specified Sy.
Acceptable compressive loads for conditions that may lead to instability if buckling occurs are 0.9 P In general, buckling due to loading in the cr.
upward vertical direction, such as hanger rod buckling, is not con-sidered as unstable.
In-plant considerations regarding other consequences of support failure such as falling and other excessive deflection shall be made when using this provision.
For expansion anchors, a minimum factor of safety of 3.0 is used when a I
complete inspection of the anchorage and surrounding concrete condition is performed. For welded anchorage, the allowable shear stress, based on the throat area, will be compared to the appropriate allowable from AISC Part lil.
I For standard component supports and hardware (e.g. U-bolts, pipe, straps, rods, pipe clamps, etc.), acceptable loads shall be:
- 1) the manufacturers recommended faulted (Level D) load rating, or
- 2) in cases where the test data is available mean ultimate capacity with a factor of safety of 2.0 (all test data must be above the calculated capacity).
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i 42094-P-003 Revision 1 May 7,1993 Page 9 of 10 5.
SEISMIC MARGIN ANALYSIS A representative portion of the Service Water system will be selected for a seismic margin analysis. The analysis will determine the system HCLPF (high confidence of low probability of failure) using the conservative deterministic failure margin (CDFM) method of Reference 4. The HCLPF I
must exceed 1.33 times the Connecticut Yankee Safe Shutdown Earthquake peak ground acceleration. This factor will give an approximate equivalency of margin between the CDFM method and the GlP method for USI A-46.
For this project, the seismic margin earthquake (SME) will be taken as the NUREG/CR-0098 (Reference 5) 84% NEP shape response spectrum for rock sites anchored at 0.23g. The ground response spectrum will be reduced to I
account for horizontal spatial variations and incoherence of input motion per Reference 4.
In-structure response spectra will be obtained by scaling the existing I
conservative design floor response spectra. The existing floor response spectra are based on a ground response spectrum with Reg. Guide 1.60 spectral shape and the structures are founded on rock. Therefore the scaling i
method of Reference 4 may be used.
First, the peak floor acceleration for the seismic margin earthquake is computed as:
- S SME/S SSE, (see Reference 4) i a
a Where A, SME s the peak acceleration at floor elevation i due to the SME, I
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Aj, ssg is the peak floor acceleration at elevation i due to the SSE, S, SME a
is the SME spectral acceleration at the dominant structural frequency and SME structural damping, and S, SSE s the SSE spectral acceleration at the i
a dominant structural frequency and SSE damping.
I The SME floor response spectrum at floor elevation iis then determined by I
using the shape of the SSE floor response spectrum at the SME damping anchored to a ZPA equal to Aj, SME. If the structural response results for j
the SSE are not available, the ZPA for the SME spectrum may be determined by multiplying the ZPA of the SSE spectrum by the ground spectral acceleration as above.
42094-P-003 Revision 1 May 7,1993 Page 10 of 10 Structural frequency variation may be accounted for by frequency shifting of I
unbroadened response spectra or by using broadened response spectra, The piping will be linearly analyzed in accordance with Reference 2 using 5%
damping per Reference 4. The analysis will use the in-structure response I
spectra in the vertical and two horizontal directions with stresses combined by SRSS. Resulting stresses in ductile piping will be multiplied by 0.8 to account for inelastic energy absorption and compared to allowables equivalent to ASME Section 111, Subsection NC, for Service Level D.
6.
REFERENCES 1.
Northeast Utilities " Piping Stress Analysis Guidelines for Seismic Qualification of Safety Related Piping at Connecticut Yankee", dated 4-29-83.
2.
Project Plan " Connecticut Yankee Atomic Power Company Seismic Upgrade Program for Service Water System, Feedwater System and Main Steam System Outside Containment", dated 11-20-92.
3.
ANSI B31.1 Power Piping Code,1973 Edition, Summer 1973 Addenda.
4.
EPRI NP-6041-SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin", Revision 1,1991.
5.
NUREG/CR-0098, " Development of Criteria for Seismic Review of I
Selected Nuclear Power Plants", Prepared for the U.S. Nuclear Regulatory Commission.
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