ML20235A514
| ML20235A514 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 06/30/1987 |
| From: | Chang K, Himler J, Tilda Liu GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20235A491 | List: |
| References | |
| NUDOCS 8707080462 | |
| Download: ML20235A514 (29) | |
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l GEORGIA POWER COMPANY V0GTLE NUCLEAR GENERATING PLANT UNIT 2 EVALUATION OF REACTOR COOLANT SYSTEM FOR STEAM GENERATOR UPPER SUPPORT REDESIGN JUNE 1987 T. H. Liu J. C. Himler K. C. Chang l l
l Approved: bOM R. B". Patel, Manager Piping Analysis & Engineering 8707080462 870701 PDR ADOCK 05000425 A PDR 23es.fc37 w as:ss u $$
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TABLE OF CONTENTS Page LIST OF TABLES iii ,
LIST OF FIGURES iv 1
I INTRODUCTION 2
11 BACKGROUND ANALYSIS 5 111 A. Mathematical Models 5 B. Loading Conc'itions 5 C. Codes and Standards 6 D. Ccmputer Programs 7 E. Structural Qualification of Piping and Supports 7 IV RESULTS AND DISCUSSIONS 8 A. Stress in Reactor Coolant Loop Piping 8
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B. Fracture Mechanics Evaluation 9 C. Support Evaluation 9 D. Primary Component Nozzle Load Conformance 10 V CONSERVATISMS 10 l
VI QUALITY ASSURANCE 11 l VII ENHANCEMENT OF KELIABILITY 11 1
VIII CONCLUSIONS 12 IX REFERENCES 12 2393a/0371s/06258710 jj
LIST OF TABLES Table 1: . Reactor Coolant Loop Piping Stresses I
Table 2: RCS Primary Equipment Support Results l
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1 LIST OF FIGURES l
1 Figure 1: Steam Generator Upper Support Snubber Arrangement Figure 2: Redesigned Steam Generator Upper Support Snubber Arrangement Figure 3: Steam Generator Support System Figure 4: Loop Piping / Support System Model Figure 5: Steam Generator Upper Support ,
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Figure 6: Steam Generator Lower Support figure 7: Reactor Coolant Pump Lateral Support 3 i
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I. INTRODUCTION This report is submitted in support of Georgia Power Company's request for FSAR changes for Vogtle Nuclear Generating Plant Unit 2. The FSAR changes would permit the reduction of several large bore hydrsulic snubbers from the steam generator unper lateral supports (figures 1, 2, and 3) as a result of excluding the dynamic effects of postulated reactor coolant loop pipe ruptures, These changes to the design basis are in compliance with limited scope General Design Criteria 4 (GDC-4) rule changes and are intended to improve support reliability and reduce occupational radiation ex;,osure.
Two hydraulic snubbers at each steam generator upper lateral support will be retained from the original design of five. Thus, a total of eight large bore hydraulic snubbers will be retained, twelve will be eliminated.
The technical basis for the FSAR changes is the use of " leak-before-break" technology for excluding from the design basic the dynamic effects of postulated pipe ruptures in primary coolant piping. Westinghouse topical reports (WCAP-10551 (Preprietary) and 10552 (Non-Proprietary), reference 1) documenting the fracture mechanics analyses were submitted under separate attachments. ]
I The purpose of this report is to demonstrate that the reactor coolant piping, components, and redesigned support configuration are able to withstand all remaining loads, including those due to the safe shutdown earthquake (SSE) and !
the limiting high energy line breaks at branch nozzles, with an acceptable margin of safety. Specifically this report will demonstrate that:
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- 1. The maximum stresses in the reactor coolant loop piping are within FSAR allowables. l l
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- 2. The reactor coolant system components and supports continue to have acceptable margins of safety.
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- 11. BACKGROUND !
A. General f 1
i The reactor coolant loop (RCL) piping of pressurized water reactors {
i (PWRs) is highly reliable, and for Westinghouse plants there is no l history of cracking failure. The Westingneuse reactor coolant system (RCS), consisting of four RCLs, has an cperating history which demonstrates its inherent stability. This includes a low susceptibility to cracking failure from the effects of corrosion (e.g., intergranular stress corrosien cracking), water hammer, or fatigue (for both low and high cycle).
1 An independent review of the design and construction practices used in Westinghouse PWR plants by Lawren:e Livermore National Laboratory (reference 2) has provided assurance that there are no deficiencies in the Westinghouse RCL design or construction which will signifi-cantly affect the probability of double ended guillotine break. The application of the " leak-before-break" technology (references 1 and 1 I
- 3) to the RCL piping eliminates the requirement to design for the
. extreme loads associated with these previously postulated pipe 1
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rupture events. This provides the opportunity to reduce the number l
of the steam generator upper support snubbers which primarily react pipe rupture loads.
Large bore Fydraulic snubbers, being active components, require periodic removal for functional testing and implementation of a seal service life program. Removal and inspection activities of large bore hydraulic snubbers will expose maintenance personnel to high levels of radiation because the snubbers are located inside the reactor containment building. The deletion of these snubbers will reduce this source of occupational exposure and facilitate mainten-ante and in-service inspections of piping and components by reducing plant congestion. Support system reliability is also increased with the removal of these active elements, because the potential large bore hydraulic snubber problems of inadvertent lockup, bleed rate variance, and hydreulic fluid leakage are reduced.
The large boro (12 in. I.D.) hydraulic snubbers on the Vogtle Unit 2 steam generator (SG) upper supports which can be eliminated are indicated in figures 1 and 2. All other ~ supports in the RCL system I l
i are not altered.
B. Steam Generator Upper Support Configurations l
The upper steam generator support consists of an octagonal ring l
girder placed around the generator shell. The girder is hung from the steam generator trunnions by four tie rods. These tie rods 2393s/0371s/062987:10 3 J
support the dead weight of the ring and aid in the vertical positioning of the girder. Laterally, the girder is connected to five hydraulic snubbers placed parallel to the hot leg on the reactor side of the steam generator. The redesigned configuration contains two hydraulic snubbers on each steam generator girder.
These snubbers, along with a strut behind the steam generator and parallel to the hot leg, restrain the steam generator for motions and loadings along the hot leg. Restraint of motions and loadings normal to the hot leg is provided by two additional struts that bear against the ring girder. These struts are attached to the secondary shield wall with embedded anchor bolt assemblies. Loads are trans-ferred from the steam generator shell to the ring girder by means of curved bearing plates bolted to the ring girder. This upper support system allows unrestrained thermal loop expansion to the final hot operating position. At this position, each strut to ring girder bearing surface is shimmed to provide proper contact, thus providing restraint to the steam generator in the operating position. A sketch of the steam generator upper supports is shown in figure 1.
The redesigned snubber arrangement is shown in figure 2. The steam generator and its combined support system are shown in figure 3.
Detailed descriptions of the remaining reactor coolant loop equipment supports (SG lower, RPV, and RCP supports) can be found in the VEGP FSAR section 5.4.14 (reference 4),
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III. ANALYSIS 1
A. Mathematical Models l
The RCL piping model consists of mass and stiffness representations for all four RCLs and the reactor vessel. Each RCL includes the primary loop piping, steam generator and reactor coolant pump. The primary equipment supports are represented by stiffness matrices.
The seismic analysis is performed by the envelope response spectra method with damping values of 2 and 4 percent for OBE and SSE, respectively.
The analysis of the RCS was performed using a four loop model with the model of one loop shown in figure 4 to obtain component and support loads and displacements. This model is identical to the one used in the Vogtle unit 1 RCS structural qualification. It is a detailed and realistic model which represents the state-of-the-art techniques in modeling.
B. Loading Conditions The RCS with the reduced snubtor support configuration was enalyzed for the following loading conditions:
o Deadweight, o Internal pressure, o Thermal expansion, 2393s 9 371s/062587 10 5
o Seismic events (OBE and SSE), and j o Postulated pipe ruptures at nozzles (pressurizer surge, occumulator, Residual Heat Removal, Main Steam and Feedwater)
For the seismic analysis, peak broadened floor response spectra for j l
two percent and four percent critical damping (OBE and SSE) were used. Responses to the three directions of earthquake loading were
- evaluated in accordance with the FSAR by combining all three 1 directional earthquakes by the square-root-sum c14the-squares (SRSS) method. The Westinghouse method of closely spaced modes combination was also used in the analysis, as defined in the FSAR.
The postulated terminal end breaks at RCt. connection in the pressurizer surge, residual heat removal (RHR), accumulator, main steam, and feedwater line nozzles were analyzed to determine the most severe loadings on the revised support configuration with the three snubbers removed. Time history forcing functions for the above five nozzle breaks were applied . .. the analytical model of f figure 4, to obtain maximum loads.
The SRSS method was used for combining pipe rupture and SSE loads.
1 C. Codes and Standards The following Codes and Standar.ds were utilized in the analysis: l o ASME Boiler and Pressure Vessel Code,Section III, Subsections NB, NF (reference 5).
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1 o Final Safety Analysis Report (FSAR) for Vogtle Nuclear Generating Plants, Units 1 and 2, Georgia Power Company (reference 4).
D. Computer Programs The RCL piping system analysis used the WESTDYN computer code (reference 6). The WESTDYN computer code has been utilized on numerous Westinghouse plants and was reviewed and approved by the NRC in 1974. The code is verified for this application and a controlled version is maintained by Westinghouse.
The modeling techniques used by Westinghouse for Vogtle Unit 2 are similar to those used for many other plants. The reliability of these techniques is assured by the Westinghouse design control process and comparison of results to other computer programs, in:1uding STARDYNE, a public domain code, and ME101, a Bechtel Er.gineering Corporation program. The comparison with STARDYNE was based on the reactor coolant loop analysis results obtained on the Surry Nuclear Units by Westinghouse and an independent Architect /
Engir.eer firm and found to be in good agreement.
E. Structural Qualification of Piping and Supports !
l The RCS structural analysis of the piping / support system was perfon.w ' .ith the redesigned steam generator upper lateral support i
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configuration with the other supports remaining the same. The results of the analyses are used to perform the stress evaluations of the piping, supports, and the connecting primary component nozzles. Revised equipment support concrete embeddment loads were compared against previous concrete ' loads and were found to be enveloped by the previous loads. Therefore the new embeddment loads are acceptable.
The stress criteria for the RCL piping and supports are presented in FSAR section 3.9.B.3 and 3.9.N.1. The faulted condition includes the SRSS combination of SSE with each of five postulated pipe ruptures at nozzles: accumulator, surge, RHR suction, mainsteam, and feedwater. The piping and support allowable stresses are obtained from the appropriate editions of the ASME Section III Code subsections NB, NF, and Division i Appendix F, respectively (reference 5).
IV. RESULTS AND DISCUSSIONS A. Stress in Reactor Ceolant Loop Piping i
Table 1 provides the level of stress in the RCL piping and the allowable stresses from the Code (reference 5), A comparison is also shown between the maximum stress in the RCL piping for the current and redesigned support configuration for controlling load i combinations. The results show that the stresces in the piping are within allowable limits.
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B. Fracture Mechanics Evaluations Westinghouse topical reports WCAP-10551 (Proprietary) and 10552 j i
(Non-Proprietary), have provided a substantial and adequate technical basis for limiting postulated design basis flaws in the Vogtle Unit 2 stainless steel primary coolant loop piping. The l
analyses have demonstrated that the probability of rupturing such 1
piping is extremely low under design basis conditions. These WCAP's j have documented the plant specific fracture mechanics study in f l
demonstrating the leak-before-break capability. j i
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With the redesigned steam generator upper lateral support configura- !
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tion, revised loads (forces and moments) in the primary coolant loop piping have been generated. The revised loads were compared with !
those loads in WCAP's 10551 and 10552, (reference 1). Based on the comparison, it is verified that the leak-before-break conclusions of WCAP's 10551 and 10552 remain valid for the redesigned support configuration.
C. Support Evaluation The faulted (LOCA and SSE) loading condition factors of safety for t
the existing and redesigned steam generator upper support configura-tions are summarized in table 2. Thefactorclfsafetyisdefinedas the ratio of the allowable support stress with respect to the actual support stress for combined loads. Loading evaluatiens performed with the redesigned support configuration demonstrate that the support stresses satisfy FSAR limits with proper factor of safety, wwem mmm 9
Figures 5, 6, and 7 illustrate the individual support element labels for the Primary Component suaport.
D. Primary Component Nozzle Load Conformance i' The RCL piping loads on the primary nozzles of steam generator, reactor coolant pump and reactor pressure vessel were evaluated.
The conformance e",:uction consisted of load component comparisons and load combina' ,on comparisons, in accordance with each of the respective Equipment-Specifications (E-spec). It was concluded that all RCL piping loads acting on the primary component nozzles were under the E-spec allowables and acceptable.
V. CONSERVATISMS This setticn addresses the adequacy of ASME Code allowables to provide sufficient margin against indirect pipe rupture from support failures. NRC funded research has concluded that the probability of indirectly-induced DEGB in RCS piping due tc earthquake is very small for all Westinghouse reactors.
l The primary coc.cponent support design uses the subsection NF of the ASME Code.
The allowable stresses were limited to 100 percent of the yield strength at the plant faulted loading condition. The more liberal ASME service level D I allowable stresses of up to 120 percent of yield were not used. In addition; 4 comparison of stresses or loads based on elastic limits is very conservative and is not a true indicator of failure or collapse load.
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Also, pipe ruptures in primary loop branch lines and in the main steam and feedwater lines were considered in conjunction with the seismic event. These postulated ruptures are major contributors to the computed stresses. As shown in table 2, the factors of safety, based on the 100 percent yield criteria, are substantial (> 1.6) for the load combination consisting of seismic SSE, deadweight, thermal, and pressure. Such large margins should mitigate concerns about seismic risk.
'1 VI. QUALITY ASSURANCE The work has been independantly reviewed as a safety-related calculation and ,
meets 10CFR50, Appendix '3, Quality Assurance requirements. The detailed results of the analyses are maintained in Westinghouse Central Files. l l
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VII. ENHANCEF.ENT OF F.ELIABILITY NUREG/CR-3718, " Reliability Analys'is of Stiff versus Flexible Piping - Status Report" (reference 8) 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 lockup, bleed rate variance, and hydraulic fluid leakage are a few of the many problems l experienced by the nuclear industry with regard to large bore hydraulic snubbers. It was further demonstrated in NUREG/CR-3718 that certain piping systems with snubbers remLved actually exhibit higher reliability than the original design. Certain large bore hydraulic snubbers proposed for elimination here act parallel to the hot leg of the RCL piping. Inadvertent lockup of these could in' duce high thermal stresses during normal plant ;
eperation. The eliminatior. of these snuobers, therefore, enhances reliability.
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VIII. CONCLUSIONS Based on the results of loading evaluations of the reactor coolant system with the redesigned support configuration the following conclusions are made:
o Piping, components, and supports are stressed within FSAR allowable limits.
o Adequate safety margins exist with respect to strength and fatigue, and RCS structural integrity will be maintained for postulated seismic and pipe rupture events.
IX. REFERENCES
- 1. WCAP-10551 (Proprietary) and WCAP-10552 (Non-Proprietary), Technical Bases for Eliminating Large. Primary Loop Pipe Rupture as a Structural Design Basis for Westinghouse Electric Corporation, 1985.
- 2. NURFG/CR-3660, UCID-19988, Volume 3, February,1985, " Probability of Pipe Failure in Reactor Coolant Loops of Westinghouse PWR Plants," Volume 3,
" Guillotine Break Indirectly Induced by Earthquakes," Lawrence Livermore National Laboratory.
- 3. 10CFR Part 50, Modification of General Design Criterion 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures, Federr.1 Register, Vol. 51, No. 70, April 11,1986, p.12502.
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- 4. Final Safety Analysis Report (FSAR), Vogt1,3 Nuclear Generating Plants, l Units 1 and 2, Georgia Power Company.
- 5. ASME Boiler and Pressure Vessel Code,Section III, Nuclear Power Plant Components, American Society of Mechanical Engineers, 1977 edition, up to and including Summer 1979 addendum (for Piping) and including Summer 1977 addendum (for Support).
- 6. " Piping Analysis Computer Codes Manual II" Westinghouse Proprietary Class 3, Westinghouse Electric Corporation, Pittsburgh, Pa.
- 7. Ward, D. A., Letter to V. Stello, ACRS Comments on the Interpretation of 10CFR Part 50, General Design Criterion 4, " Environmental and Missile i Design Bases," December 17, 1986.
- 8. Lu, S. C. and Chou, C. K. " Reliability Analysis of Stiff vs. Flexible Piping, NUREG/CR-3718, Lawrence Livermore National Laboratory, Livermore, California, 1984.
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I TABLE 1 MAXIMUM REACTOR COOLANT LOOP PIPING STRESSES Current Redesigned ASME Code Configuration Configuration Allowable j (1) Stress Stress Stress <
ASME Code Equation M (ksi) (ksi) (ksi) )!
3650 (9) Design HL 28.9 29.7 31.5 XL 28.9 29,7 31.5 CL 28.9 23.6 31.5 )
3650 (9) faulted HL 56.6* 45.9 56.7 XL 56.6* 44.5 56.7 CL 56.6* 47.5 56.7 3650 (13) HL 55.9 56.4 57.5 XL 58.3 59.1 59.4 CL 48.4 48.1 59.4 l
3650 (usage factor) HL 0.5 0.5 1.0 j XL 0.5 0.5 1.0 CL 0.4 0.4 1.0 l
NOTES:
(1) HL - Hot Leg, XL - Crossover leg, CL - Cold leg I
- Ccnservative stress values for combination of highest stresses from deadweight, SSE, LOCA, and jet loads regardless of location in the loop.
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5.4.14 COMPGNENT S'vPORTS k 3
cV 5.4.14.1 Desian Bases Component supports allow unrestrained lateral thermal movenient 4 of the loop during plant operation and provide res_raint to the $
loops and components during accident and seismic condi tions.
The loading combinations and design stress limits are discussed y j in paragraph 3.9.B.3.<a) Support design is in accordance with gl22 7 b
the American Society of Mechanical Engineers (ASME) Code, Section 7 m Ill, Subsection NF. The design maintains the integrity of the f 0 <
RCS boundary for normal, seismic, and accident conditions and satisfies the requirements of the piping code. The results of Q4k piping and supports stress evaluation are presented in secticn a{g '
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$Qm ConformancewithRegulatoryGuides1.124and1.130isdiscussed[g in section 1.9. g i k u 5.4.14.2 Description 3 The. support structures are welded structural steel sections. age q g gq s l l
Linear-type structures (tension and compression struts, o o Wg l columns, and beams) are used in all cases except for the kt t Qg reactor vessel supports, which are plate-type structures. o -
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Attachments to the supported equipment are nonn tegral type 4 that are bolted to or bear against the compone m ports-to-concrete attachments are either anchc The sup-bults or em-qe ( g b} l 2 4 bedded fabricated assemblies. L < tu e$ 4 2t d.
The supports permit virtually unrestrained thermal growth of % *2 c y%
- 2 g the supported systems but restrain vertical, lateral, tional movement resulting from seismic and pipe break loadings. 7and rota- v This is accomplished using spherical bushings in the columns o &a for vertical support and girders, bumper pedestals, hydraulic d~ to snubbers, and tie rods for lateral support. ,n j$k Because of manufacturing and construction tolerances, ample M p 4 ([
-( q adjustment in the support structures is provided to ensure 2 , s proper erection alignment and fit-up. This is accomplished by $
shimming or grouting at the supports-to-concrete interface and w oo jE$g{h 3
by shimming at the supports-to-equipment interface.
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- a. Reference 1 provides the original criteria for pcs.tulating breaks in the reactor coolant loop. The basis for eliminating eight of these postulated large pipe breaks in the react 22 s coolant loop io providcd in reference W h ever, this destqn I
of-the-reau-tor--soe+ ant-systemHMHomponent-supp or4s--r-e mains trnehenged, -
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- i. VEGP-FSAR-5 5.4.14.2.1 Reactor Pressure Vessel f Supports for the reactor vessel (figure 5.4.14-1) are individ-ual, air-cooled, rectangular box structures beneath the vessel nozzles bolted to the primary shield wall concrete. Each box (g structure consists cf a horizontal top plate that receives loads from the reactor vessel shoe, a hor 2tontal bottom' plate k that transfers the loads to the primary shield wall concrete, y and connecting vertical plates. The supports are air-cooled to T maintain the supporting concrete temperature within acceptable '
levels.
5.4.14.2.2 Steam Generator y As shown in figure 5.4.14-2, the steam generator supports con-sist of the following elements: -
A. Vertical Support T Four individual columns provide vertical support for each steam generator. These are bolted at the top to the steam generator and at the bottom to the concrete
'j k structure. Spherical ball bushings at the top and g y
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bottom of each column allow unrestrained lateral IMh h" travement of the steam generator during heatup and cooldown. The column base design permits both hori- (* h zental and vertical adjustment of the steam generator %
for erection and adjustment of the system.
B. Lower Lateral Support 2*
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Lateral support is provided at the generator tube k sheet by fabricated steef girders and struts. These are bolted to the compartment walls and include kg h (
bumpers that bear against the steam generator but permit unrestrained movement of the steam generator dh g
duringchanges in system temperature. Stresses in the R,j gg .
1 beams caused by wall displacements during compartment N l pressurization are considered in the design. #g3q i 3 g C. Upper Lateral Support puc, 3 Upper lateral support of/the steam generator is pro-f, vided by a builtup ring / plate girder at the operating
- M eck 3 Two-way acting ^ snubbers restrain sudden seismic kIQk
@ or blowdown-induced motion but permit the yorpmal thermal movement of the steam generator./T Movement 'J perpendicular to the thermal growth direction of the steam generator ic prevented by struts.
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