ML20213D760
| ML20213D760 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 10/31/1986 |
| From: | VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | |
| Shared Package | |
| ML19292G230 | List: |
| References | |
| NUDOCS 8611120219 | |
| Download: ML20213D760 (30) | |
Text
- .. ..
ATTACHMENT 3 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA UNITS 1 AND 2 REACTOR COOLANT SYSTEM LOADS AND COMPONENT SUPPORT MARGINS EVALUATION FOR ELIMINATION OF REACTOR COOLANT SYSTEM, MAIN LOOP PIPE BREAK PROTECTIVE DEVICES
[
f OCTOBER 1986 f
8611120219 861106 PDR ADOCK 05000338 P PDR 60-KKD-4686S-1
TABLE OF CONTENTS I.- INTRODUCTION II. BACKGROUND III. ANALYSIS A. Mathematical Model B. Loading Conditions C. Codes and Standards D. Computer Programs IV. RESULTS AND DISCUSSIONS A. Stress in Piping B. Fracture Mechanics Evaluation C. Component Support Evaluation V. ADDITIONAL CONSERVATISM VI. INDEPENDENT VERIFICATION VII. QUALITY ASSURANCE VIII. ENHANCEMENT OF RELIABILITY IX. CONCLUSIONS X. - REFERENCES LIST OF TABLES Table 1: Reactor Coolant Loop Piping Stresses Table 2: Reactor Coolant Loop Natural Frequencies and Modes Table 3: Factors of Safety for Component Supports LIST OF FIGURES Figure 1: Steam Generator and RC Pump Supports Figure 2: Steam Generator and RC Pump Lower Support Figure 3: Westinghouse Mathematical Model Figure 4: Stone & Webster Mathematical Model Figure 5: SWEC/ Westinghouse Interface 60-KKD-4686S-2
F, I. INTRODUCTION This report is submitted in support of Virginia Electric and Power Company's request in compliance with General Design Criteria 4 (GDC-4)
(Reference 1) for a licensing amendment to license NPF-4 and NPF-7 for North Anna Units 1 & 2 to permit redesign of primary coolant loop heavy component support to reflect the exclusion of dynamic effects resulting from postulated pipe ruptures in the primary coolant loop piping. This redesign would allow elimination of 18 large bore snubbers from each unit which are required only for pipe rupture loadings on the reactor coolant system piping, components and supports. During this redesign six other large bore snubbers per unit would be replaced by rigid struts, taking into consideration the minimal thermal movement at those snubbers.
The technical basis for the licensing amendment is based upon use of
" leak-before-break" technology for excluding from the design basis the dynamic effects of postulated pipe ruptures in primary coolant piping as allowed by the current GDC-4 rule. Westinghouse reports (WCAP
-11163/11164) documenting the fracture mechanics analyses in support of
" leak-before-break" for North Anna Units 1 and 2 are submitted as a separate attachment.
The purpose of this report is to demonstrate that the reactor coolant piping, components and supports with the redesigned support configuration are able to withstand all remaining loads, including those due to safe shutdown earthquake, with an acceptable margin of safety. Specifically, 115-BSD-2213N-a-2
- 1. The maximum stresses in the primary loop piping are still within allowables.
- 2. The cumulative usage factor is within acceptable limits in primary loop piping and components.
- 3. The reactor coolant system equipment and supports continue to have acceptable margins of safety under all remaining licensed loading conditions.
II. BACKGROUND The primary loop piping of Pressurized Water Reactors (PWRs) is highly reliable and for Westinghouse plants (including North Anna Units 1 and
- 2) there is no history of cracking failure. The Westinghouse Reactor Coolant System (RCS) primary loop has an operating history which demonstrates its inherent stability. This includes a low susceptibility to cracking failure from the effects of corrosion (e.g., intergranular stress corrosion cracking), water hammer, or fatigue (for both low and high cycle). This operating history totals over 400 reactor years, including five plants each having over 15 years of operation and 15 other plants each with over 10 years of operation.
The application of the leak-before-break approach (References 8 and 9) to prevent ruptures of the primary coolant loop piping eliminates the requirement to design for the extreme loads associated with these 60-KKD-4686S-4
. - . . - _ , . ~ . ~ . -.
previously postulated pipe rupture events. This provides the opportunity to eliminate selected primary component support snubbers which principally carry pipe rupture loads.
Large bore snubbers, being active components, require periodic removal for functional testing and implementation of a seal service life program. Removal / inspection activities of large bore snubbers have exposed maintenance personnel to high radiation because the snubbers are located in the reactor containment cubicles. The deletion of these snubbers will eliminate this source of occupational exposure and facilitate maintenance and in-service inspections of piping and components by reducing plant congestion.
1 Support system reliability is also increased with the removal of these active elements. Inadvertent lockup, bleed rate variance, and hydraulic fluid leakage are possible large bore snubbers problems that are eliminated.
The function of each primary loop support snubber has been reviewed to
- determine those which may be eliminated. The object of this review
- was to maximize snubber elimination allowed by the elimination of dynamic effects of main reactor coolant loop breaks while having I
minimal effect on the design margins for other loads. Except for the application of " leak-before-break" for elimination of dynamic effects of main reactor coolant loop breaks, and the combination of remaining pipe rupture loads with seismic loads using the square root of the sum of the squares (SRSS) combination, other licensing basis requirements i are maintained.
60-KKD-4686S-5
e Snubbers on the North Anna Steam Generator.(SG) and Reactor-Coolant F
Pump (RCP) supports which'may be eliminated by use of
" leak-before-break" criteria are indicated in Figures 1 and 2. It is clear from the orientation of snubbers 5, 6, 7.and 8 that these snubbers were intended-primarily to carry pipe rupture loads. Their stiffness is insignificant when compared to the unbroken loop piping.
, The snubbers to be eliminated, as noted in Figures 1 and 2 are described below:
Both large bore (14" I.D.) snubbers (Nos. 5 and 6) acting parallel to the hot leg in each of the three lower-SG supports.
Both large bore (10" I.D.) snubbers (Nos. 7 and 8) acting parallel to the cold leg in each of the three RCP supports.
Two large bore (10" I.D. & 14" I.D.) snubbers (Nos. 9 and 12) acting between the RCp support and the SG lower support in each cubicle.
Also, during this redesign the two large bore (14" I.D.) snubbers (Nos. 2 and 4) located at each SG upper support and acting in a direction perpendicular to the hot leg would be replaced by rigid struts. The exiting snubbers at those locations have minimal thermal movement. It has been determined that these snubbers can be replaced by struts even with the current license condition. These struts will serve as lateral supports and permit normal steam generator movement in the direction radial from the reactor during thermal movement of the reactor coolant loop.
60-KKD-4686S-6
The snubbers to be retained, as noted in Figures 1 and 2, are described below:
Two large bore (10" I.D.) snubbers (Nos. 10 and 11) acting between the RCP support and SG lower support in each cubicle.
Two large bore (14" I.D.) snubbers (Nos.1 and 3) located at SG upper support and acting in a direction parallel to _the hot leg.
Loading evaluations performed with the revised support configuration establish that the piping, components and supports are stressed within UFSAR acceptable limits.
III. ANALYSIS i
A. Mathematical Models Two -independent analyses of the primary RC loop were performed for this work. Westinghouse Electric Corporation performed analyses using the model of Figure 3 primarily to obtain RCL equipment and piping stresses. Stone & Webster Engineering Corporation (SWEC) performed analyses using the model of Figure 4 principally to obtain component support loads. This division of analytical responsibility between the two organizations is similar to the original division of design responsibility. Both analytical models were revisions to existing models and 60-KKD-4686S-7
incorporated changes due to the proposed modifications to the supports.-
B. Loading Conditions The reactor coolant system with the revised support configurations were analyzed for the following loading conditions:
Deadweight Internal-pressure Thermal expansion Thermal transients Seismic events (0BE and DBE), and Dynamic effects of postulated pipe ruptures in other systems as specified in the UFSAR (pressurizer surge, accumulator, residual heat removal, main steam, and main feedwater lines).
These loading conditions cover all UFSAR specified loadings other than the dynamic effects of postulated pipe ruptures in the main reactor coolant piping.
60-KKD-4686S-8 L
No other hydraulic transient loading was considered as significant.
For seismic analysis, peak spread amplified response spectra (ARS) for 0.5 percent equipment damping (0BE) and 1 percent equipment damping (DBE) were used in accordance with the UFSAR.
Responses to the three directions of earthquake loading were combined by SRSS. The combination of closely spaced modes conformed to the "10 percent method" of NRC Regulatory Guide 1.92, Rev. 1.
The Squarc root of sum of squares was used for combining pipe rupture and DBE loads.
C. Codes and Standards The following Codes and Standards were utilized in the analysis:
Power piping, USAS B31.7-1969 and Addendum through 1970 (Reference 5). This is the original' code of ,ecord to which the plant was constructed.
Updated Final Safety Analysis Report (UFSAR), North Anna Power Station Units 1 and 2, Virginia Power. Allowable stresses currently documented were used for requalification.
60-KKD-46865-9
i ASME Boiler and Pressure Vessel Code, Section III-1971, Nuclear
~
Power. Plant components .(Reference 6), was used for the~ design and construction of the equipments.
Di Computer Programs-
~
The Westinghouse analysis used the WESTDYN computer code (Reference 2). The WESTDYN compu'ter code bas been utilized on numerous. Westinghouse plants and was'raviewed and found
. acceptable by the NRC in 1974. ' Component support stiffness matrices were supplied by SWEC.and are identical to matrices used in earlier analyses, except for stiffnesses representing the eliminated snubbers..
The SWEC analysis used the STARDYNE computer code (Reference 3).
STARDYNE is a public domain computer program and is recognized as a Category I computer program suitable for nuclear work. The following modules of STARDYNE, Version 3 Level H, were-used:
STAR (Static and Modal Extraction)
DYNRE4 (Seismic Response Spectrum)
DYNRE6 (Time History Transient Analysis) - only used for evaluating pipe rupture loadings, 60-KKD-4686S-10
r-E STARDYNE is maintained land monitored with SWEC's Quality Control procedures with respect to any program errors which are encountered through industry usage.
IV. RESULTS AND DISCUSSIONS A. Stress in Reactor Coolant Loop Piping The Table 1 provides a comparison between the level of stress in the piping and the UFSAR allowable stresses. A comparison is also shown between the maximum stress in the reactor coolant loop piping in the existing and proposed support configurations for controlling combined loading conditions. The results clearly show that stresses in piping are well within allowables. The location o/ maximum stresses for seismic loading conditions virtually remain unchanged. The cumulative usage factors indicate an adequate margin of safety against fatigue.
B. Fracture Mechanics Evaluation Forces and moments developed at several salient locations of the reactor coolant loop pipings for different loading conditions and proposed support configuration were used in a fracture mechanics evaluation of those locations.
Westinghouse report WCAP-11163/11164 which are included as separate attachment justify the elimination of RCS primary loop pipe breaks for the North Anna Units 1 and 2 as follows:
115-BSD-2213N-a-3
- 1. Stress corrosion cracking .is precluded by use of fracture resistant materials in the piping system and. controls on reactor coolant chemistry,' temperature,- pressure, and flow during normal operation.
- 2. Water hammer should not occur in the RCS piping because of system design, testing, and operational considerations.
- 3. The effects of low and high cycle fatigue on the integrity of the primary piping are negligible.
- 4. A large margin exists between the leak rate of the reference flaw and the criteria of Reg. Guide 1.45.
- 5. Ample margin exists between the. reference flaw chosen for leak detectability and the " critical" flaw.
- 6. Ample margin exists in the material properties used to demonstrate end-of-service life (relative to aging) stability of the reference flaw.
The reference flaw will be stable throughout reactor life because of the ample margin in items 4, 5, and 6 above and will leak at a detectable rate which will assure a safe plant shutdown. A detailed review of leakage detection capabilities at North Anna 1
& 2 are included as Attachment 2.
i l
! 60-KKD-4686S-12
-Based on the above, it is concluded that postulation of primary loop pipe breaks need not be considered in the structural design basis of the North Anna Units 1 and 2.
C. Component Support Evaluation Using the analytical model of Figure 4, SWEC has evaluated the proposed support system configuration. As indicated by Table 2, the frequencies of significant vibrational modes are virtually unchanged. The loads on primary equipment and supports continue to be low.
The postulated terminal and intermediate breaks in the pressurizer surge, residual heat removal, accumulator, main steam and main feedwater lines were reviewed by SWEC to determine those breaks which would cause the most severe loadings on the revised support configuration with snubbers removed. Time history forcing functions were applied to the analytical model of Figure 4 representing these potentially limiting breaks, to obtain maximum loads with the revised support configuration. These loads were combined by SRSS with seismic DBE loads and then summed with deadweight and pressure loads. In all evaluated cases, the support loads are within UFSAR and code allowables.
Table 3 provides Factors of Safety for the existing and proposed support configurations under combined Deadweight, Pressure, Thermal, and SRSS (DBE and Pipe Pupture) loading where:
60-KKD-4686S-13
Factors of Safety , Allowable Load Calculated Combined Load .
.7 f
For the component interfaces, the Allowable Load is tak'en frein
[ lthe conservative Westinghouse Equipment Specification and, ,
thersfore", contains' additional conservatism.
All' support components exhibit Factors of Safety of over 1.3 with the exception. of the upper Steam Generator supports. The snubber
' and strut load at the upper steam generator supports are governed by postulated ruptures of the Main Steam lines. Because the l
' ruptures were postulated to give maximum design loads, the lowest
- Factor of Safety is 1.2 for those snubbers. The proposed change.
. to eliminate the lower snubbers has a negligible effect on the loads on the. remaining upper steam generator snubbers, which :
continue to have a Factor of Safety of 1.2. Similarly, there is e
l no increase in loads on the upper steam generator strut locations due to the elimination of snubbers at the lower support. The 4- struts carry insignificant thermal loads, since the struts are oriented in a direction perpendicular to the direction of hot leg where the expected thermal effect is minimal.
i V. ADDITIONAL CONSERVATISM a
j The analyses performed are in accordance with the existing licensing '
basis, except for the SRSS combination of pipe rupture loads with
! seismic loads. The factors of safety, quoted in Tables 1 and 3, are 115-BSD-2213N-a-4 v -< =++-m + y -,,wy-,we,e-ww.-- -,,,-w,* --*-y,,--,. r, . ,e-+,,-merm-.- - - - - + - + * ---v-*--+tw,----Mm--+,e-,.ee ret'- w+*e-w*wr--www
based upon criteria more conservative than the current industry practice. Additional conservatisms include:
The use of low equipment damping ~(0.5 percent for OBE,1 percent for DBE) compared to higher values recommended in Regulatory Guide 1.61 (Reference 7).
Comparison of stresses to minimum code-specified material allowables at operating temperature (References 5 and 6), which already include a safety factor, and Comparisons based on elastic limits which are not a true indicator of failure.
VI. INDEPENDENT VERIFICATION As discussed previously, two essentially independent analyses were performed by Westinghouse and Stone and Webster Engineering Corporation (SWEC). Both completely modeled a single primary loop.
The results of both analyses at suoport-to-component interface points were reviewed and found to be in close agreement. Reactor Coolant loop natural frequencies and modes listed in Table 2 show good agreement in results obtained by Westinghouse and SWEC.
lhe interfaces between Westinghouse and SWEC for this work have been t
carefully monitored. Interface details are provided in Figure 5.
60-KKD-46865-15 i.
1
, . . _ y . - . - - . - _ _ . _ . . .,..,__,.m. , ,_ .__-_._,,py._ ,m
The two analytical models prepared by Westinghouse and Stone & Webster (Figures 3 and 4, respectively), were revisions to the mathematical models prepared for the original design analysis of the primary coolant loop piping and equipment supports.
These models had been developed based on-the Unit 1 design drawings.
The corresponding Unit 2 drawings are sufficiently similar to Unit I to justify use of the same model. The Reactor Coolant System (RCS) piping is shop-fabricated " fitting-to-fitting" construction with strict tolerances, due to the relatively short lengths of pipe and the requirement to accurately locate the branch piping. In a very real sense, the plant is designed around the primary loop, and the types of interferences which cause minor routing and support relocations in other piping systems, which can lead to as-built differences, are not pennitted in the RCS loop piping.
The primary equipment supports were shop fabricated to strict tolerances. The components and supports were then accurately positioned in accordance with detailed installation specifications, and finally, the loop piping was installed between the components.
The RCS primary piping is supported only by the RCS primary equipment.
Therefore, there is no concern about incorrect support location.
The computer models prepared by Westinghouse and SWEC were based on the original models. As part of the current re-analysis, these models were reviewed and modified to reflect the proposed snubber elimination and substitution of some rigids. Personnel involved in these reviews 60-KKD-4686S-16
jW -
7 y -
differed from those involved in generating the original models. .The-
~
3 calculations, including modifications to the analytical models,.were.
independently reviewed.
+
= ..,'
[ The sbpport configuration is essentially the same as the original configuration, except for removal of the snubbers. -The Reactor
~ '
Coolant Pump support stiffness matrix without snubbers provided to Westinghc.use by SWEC'is essentially the same'as previously provided T '- for the deadweight and thermal cases (i.e. with snubbers inactive).
The stiffness matrix for the steam generator support reflects
,k - eliminating some of the -lower support snubbers, and for substituting -
two upper steam generator support snubbers by struts.
,e c
'[<
Comparison _of interface loads calculated by two models was performed to ensure the results of the two models were consistent. The signiff_ cant interface loads were found to be _in good agreement. The close comparison of natural frequencies and modes in both model (Table 2)addsfurthertoindependentverificationofanalysis.
VII. . QUALITY ASSURANCE Except.for elimination of dynamic effects due to pipe rupture of the primary RCS piping and SRSS combination of other pipe rupture loads with seismic loads, analyses are in conformance with the existing licensing basis (Reference 1), both with respect to design criteria and the control of the engineering process. The work has been independently reviewed as safety related calculations and meets 10 CFR 60-KKD-46865-17
50 Appendix B Quality Assurance requirements. The results of the analyses are maintained in Project Document Control.
VIII. ENHANCEMENT OF RELIABILITY NUREG/CR-3718, " Reliability Analysis of Stiff versus Flexible Piping-Status Report" (Reference 10), established that piping designs using snubbers as support devices may not exhibit the intended reliability because the snubbers may fail to perform the desired function. Inadvertent lock-up, bleed rate variance, and hydraulic fluid leakage are a few of the many problems experienced by the nuclear industry with regard to large bore snubbers. It was further demonstrated in the NUREG/CR-3718 that certain piping systems with snubbers removed actually exhibit higher reliability than do those of the original design. The large bore snubbers proposed for elimination here are parallel to both the cold leg and the hot leg of the reactor coolant loop piping. Inadvertent lock-up of these could induce high thermal stresses during normal plant operation. The elimination of these snubbers therefore enhances reliability.
The revised support configuration will eliminate snubbers in high radiation areas and the more inaccessible areas. Per NUREG/CR-4279 (Reference 11) reduction in number of snubbers results in enhancement of reliability. The large bore snubbers to be retained in the main reactor coolant system will be in more accessible area; and, therefore, can be maintained more readily so as to increase their reliability. The snubbers retained can be equipped with seals with 60-KKD-4686S-18
- y l
longer service life, test-in-place capability, and valves with superior lock,-up and bleed function to further enhance reliability.
Therefore, the revised support configuration would result in improved overall reliability of the RCS support system.
IX. CONCLUSIONS ,
Based on the results of loading evaluations of the reactor coolant system with proposed support configuration the following conclusions are made.
- Piping, components and support are stressed within UFSAR allowable limits.
Adequate safety margins exist with respect to strength and fatigue, and structural integrity will be maintained du' ring seismic events. '
X. REFERENCES
- 1. Updated Final Safety Analysis Report (UFSAR), North Anna Power Station Units 1 and 2, Virginia Power.
- 2. WESTDYN, Westinghouse Electric Corporation.
s
- 3. STARDYNE, Version 3, Level H System Development Corporation, February 1, 1984.
60-KKD-4686S-19 8
i
- 4. Regulatory Guide 1.92, Rev.1, Combining Modal Responses and Spatial Components in Seismic Response Andysis, U.S. Nuclear
. Regulatory Commission, February 1976.
- 5. USAS B31.7, Power Piping, American Society of Mechanical Engineers, 1969 and Addenda through 1970.
- 6. ASME Boiler and Pressure Vessel Code,Section III,1971, Nuclear Power Plant Components, American Society of Mechanical Engineers.
- 7. Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants, U.S. Nuclear Regulatory Commission, October 1973.
- 8. WCAP-11163/11164, Technical Bases for Eliminating Large Primary Loop Pipe Rupture As A Structural Design Basis for North Anna Units 1 and 2, Westinghouse Electric corporation, June 1986.
- 9. USNRC Generic Letter 84-04, Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops, U. S. Nuclear Regulatory Commission, February 1,1984.
- 10. Lu, S.C. and C.K. Chou, Reliability Analysis of Stiff vs.
Flexible Piping, NUREG/CR-3718, Lawrence Livermore National Laboratory, Livermore California 1984.
60-KKD-4686S-20
- 11. Bush, S.H., P.G. Heasier, R.E. Dodge, Aging and Service Wear of
_ Hydraulic and Mechanical Snubbers Used on Safety-Related Piping and Components of Nuclear Power Plants, NUREG/CR-4279, Pacific
' Northwest Laboratory, February,1986.
h 5
4 60-KKD-4686S-21
TABLE 1 REACTOR COOLANT LOOP PIPING STRESSES ASME Code Location Stresses Code allow- Stress Ratio Equation Node (ksi) able (ksi) to allowable 9-design HL-1112 18.5 23.2) 26.7 0.693 0.869 XL-142 18.3 22.7) 26.7 0.685 0.850 CL-177 19.3 21.8) 26.7 0.723 0.816 9-faulted HL-1112 40.09 46.05) 53.4 0.751 0.862 XL-140 25.23 41.8 53.4 0.472 0.783 CL-177 49.69 48.0 53.4 0.931 0.899 12 HL-1112 19.8 18.2) 54.06 0.366 (0.337)
XL-157 10.09 10.6 55.74 0.181 0.190 CL-1175 6.09 7.4 55.74 0.109 0.133 13* HL-103 39.8 54.06 0.736 0.736 XL-139 37.8 55.74 0.678 0.678 CL-179 (38.1) 55.74 0.684(0.684)
CUMULATIVE USAGE FACTORS
- Location Cum Usage Node Factor Allowable HL-103 0.6576 (0.6576) 1.0 XL-1149 0.0500 0.0500) 1.0 CL-179 0.0990 0.0990) 1.0
- Same values as existing stress report since the conservative values were used in the analysis. Forces and moments in the proposed support configuration are
. bounded by the values used in existing configuration.
Note: 1 Numbers inside parentheses refer to existing support configuration 2 HL - Hot leg of reactor coolant loop 3 XL - Cross over leg of reactor coolant loop 4 CL - Cold leg of reactor coolant loop 5 For node numbers refer to Figure 3 Equation 9 - design: Includes stresses due to pressure, deadweight and OBE.
Equation 9 - faulted: Includes stresses due to pressure, deadweight and SRSS of (SSE and governing pipe rupture).
60-KKD-4686S-22
TABLE 2 NORTH ANNA R.C.L. iiATURAL FREQUENCIES AND MODES Frequency (Hz)
Proposed Design Existing Design- (See Note 2) Description of Mode 6.419 6.416 S.G. rocking parallel to hot leg (6.19) 6.458 6.363 S.G. rocking perpendicular to hot leg (6.77) 9.723 8.505 RCP rocking perpendicular to cold leg (8.37) 14.12 13.87 RCP rocking parallel to cold leg (12.44) 15.84 14.50 RCP rocking with CL and stop valve perpendicular (12.75) to cold leg.
16.85 16.25 S.G, crossover leg, hot leg, stop valve vertical.
(15.63) RCP rocking perpendicular to hot leg.
19.04 17.72 Steam Generator rocking parallel to hot leg.
(15.02) 21.72 20.60 Crossover leg perpendicular to hot leg.
(18.48)
Notes:
(1) The above modes account for more than 97 percent of the resultant generalized displacement.
(2) Numbers inside parentheses were determined by Westinghouse using model in Figure 3. Numbers without parentheses were generated from SWEC model Figure 4.
60-KKD-4686S-23
TABLE 3 FACTORS OF SAFETY FOR COMPONENT SUPPORTS UNDER COMBINED LOADS
- MINIMUM FACTOR OF SAFETY COMP 0NENT EXISTING DESIGN PROPOSED DESIGN Steam Generator Shell Bending 3.5 3.1 Steam Generator. Upper Support Most Critical Member 2.5 2.4 Struts (Substituted for Existing Snubbers)' -
1.2 Snubbers 1.2 1.2 Steam Generator Lower Support Most Critical Member 1.4 1.5 Steam Generator Foot Vertical Force 3.4 3.3 Tangential Force 5.0 4.5 RC Pump Support Most Critical Member 1.4 1.3 RC Pump Foot Vertical Force 1.3 1.4 Tangential Force 2.9 3.1 Snubbers Remaining Between RCP & SG Supports 3.7 3.4
- \ Deadweight l+ l Pressure l+ Thermal l+ SRSS of (SSE and remaining postulated pipe rupture)\
115-BSD-2213N-a-5
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