ML20141M357
| ML20141M357 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 05/31/1992 |
| From: | Baker T, Chan J, Tilda Liu WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20141M355 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM NUDOCS 9208110389 | |
| Download: ML20141M357 (25) | |
Text
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i Westinghsuse Proprietary Class 3 PRESSURIZER SAFETY LINE PIPING AND SUPPORT EVALUATION UNDER SAFETY VALVE DISCHARGE LOADING 1
SOUTHERN NUCLEAR OPERATING COMPANY-
- J. M. FARLEY UNIT 1 AND UNIT 2 T. H. Liu J. C. Himler Verified by:
N '
' Verified by: ~I kl . b -
J. ICChart . T. Daker -
z Approved by:-
- R. B. Patel, Manager.
System Structural Analysis & Development 4
i May 1992 L WPF12%J/052192:10 9208110389'920807
PDR -ADOCK=05000348 P. PDR-
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, 4 . t , r, Westinghouse Proprietary Class 3 TABLE OF CONTENTS Section Thic Eags 1.0 Introduction 1 I
1.1 General 1 1.2 Farley Plant Units 1 and 2 2 -I 2.0 Arialytical Modeling and Approaches 3 l 2.1 Thermal Hydraulic Modeling 3 2.2 Structural Modeling 5 2.3 Elastic / Plastic Analysis Methods 5 3.0 Piping Component and Support Evaluation 6 3.1 Piping Co.nponent/ Support Systems Evaluation Criteria 6 3.2 Piping Compoa..:t Evaluation Results 7 33 Support Component Evaluation Results- 9 4.0 Summary and Conclusions 10 5.0 References 12 Tables and Figures 14 24-WPF1276J/052192:10 [j-
Westinghouse Proprietary Class 3
1.0 INTRODUCTION
1.1 General The Pressurizer Safety and Relief Valve (PSARV) discharge piping system for pressurizer water reactors, located on the top of the pressurizer, provides overpressure protection for the reactor coolant system. A water seal is maintained upstream of each pressurizer safety and re!!ef valve to prevent a steam interface at the valve seat. This water seal practically eliminates the possibility of valve leakage. However, with this arrangement, the water s'fug, driven by high system pressure upon actuation of the valves, generates severe hydraulic shock loads on the piping and supports.
Under NUREG 073791, Section ll.D.1, "Performar'ce Testing of BWR and PWR Relief and Safety Valves," all operating plant licensees and applicants are required to conduct testing to qualify the reactor coolant system relief and safety valves under expected operating conditions for design basis transients and accidems. In addition to the qualification of valves, the functionability and structural integrity of the as built discharge piping and supports must also be demonstrated on a plant specific basis.
In response to these requirements, a prcgram for the performance testing of PWR safety and relief valves was formulated by EPRl:21. The primary objective of the Test Program was to provide full scale test data confirming that functionability of the reactor coclant-system power operated relief valves and safety valves are capable of performing their design fimetion for expected operating and accident co:.l.*: ons. The second objective of
. the program was to obtain sufficient piping thermal hydraulic load data to validate models utilized for plant unique analysis of PSARV discharge piping systems. Based on the results of.the aforementioned EPRI Safety and Relief Valve Test Program, additional thermal hydraulic analyses _were required to adequately define the loads on the piping system due to valve actuation, a
i WPF1276J/052192:10 1 'l
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' Westinghouse Proprietary Ctess 3 1.2 Farlev PlanLUnits 1 and.2 In response to NRC letter of September 29,1981 and to the requirements of NUREG-0737, '.iern II.D.1, Alabama Power Cornpany submitted letter rc.ponsesPIlilt o NRC or.
April 2,1982 and July 1,1982. In both letters, Alabama Fower Company addressed the capability of relief and safety valve and piping issues by committing to complete further analysis of the downstream loads due to valve actuation based on the results of EPRI test program. Subsequent to those ;ubmittals, on November 4.1982, Alabama Power Company informed the NRC that the analysis had been completed DI . A Westinghouse report was attached to that November 4th letter. It was indicated that the evaluations performed by Westinghouse were based on cold seal discharge which is the design basis for Farley Nuclear Plant. It was concluded that no overstress occurred in piping subsequent to actuation of the pcwor operated relief valves. However, a potentia'.
overstressed region in the piping downstream of the safety valves was identified subsequent to safety valve discharge. These results are based on the conservative postulation that all three safety valves simultuneously discharge cold loop seals. In a December 16,1986 safety evaluation report (SER),MI he t NRC expressed concern regarding the potential impact of the operability of the safety valves due to the over-stress condition in the pipe. The SER postulatea that the over stressed pipe may deform rather than ruptere, thus affecting the safety valves overpressure protection capacity.
In response to the SER, Alabama Power Company made a commitment to NRC in a letter dated February 5,1987 to provide a schedule for resolution of NRC concernsFl. In a letter, dated September 16,1988, the NRC was informed of Alabama Power Company's goal of raising the temperature of 'he water in the loop seal piping!'l. To achieve this goal, modifications to piping insulation were necessary to ensure sufficient heat is conducted to the loop seal water. Upon the completion of tnese modifications, at-power temperature measurements of the loop seal piping would be made. These modifications, in conjunction with the inspection, test and maintenance procedures were )
considered to be the resolution of all remaining issues associated with NUREG 0737, Item II.D.1.
WPF1276J/052192:10 2 i
Westinghouse Proprietary Clou 3 In 1991, the above stated temperature measurementt were obtained for both unhsM.
'The modified insulation did serve the intended purpose to raise the ternperature of the i water in the loop seal piping, and therefore, reduce the severity of the hydrsulic sho:k loads from water slug discharge on the piping and supports. This report is being prepared to discuss the analysis and results from these new loads in the piping and support system.
2.0 ANALYTICAL MODELING AND APPROACHES 2.I Thermal Hvdraulic Modeling
%e safety valve dis:harge loads were calculated for the fluid transient condition that will produce the most severe loading on the piping system. His occurs durir:g a high pressure steam transient where steam from the pressurizer forces the water in the water seal ?:ough the safety valve down the piping system to the relief tank. Forcing functions are normally generated for hot or cold ioop seals depending on the temperature in the ,
loop seal. The hot and cold loop seal conditions for Farley plants are consistent with the hot and cold loop seal conditions defined in 1982 EPRI tests. De general arrangement of a safety valve loop seal is shown in Figure 1.- Thr.rmal hydraulic analysis for the Farley pressurizer safety valve (PSARV) system were eriginally analyzed in 1982 for both the hot and cold loop seal conditions. De hydraulic forces 3cnerated when the safety valves open are much higher for the cold loop seal condition compared to those forces from the hot loop seal condition. To redu e the loan from cold loop seal condI2 ion, modification to piping insulation was necessary to ensure 6allicient heat was conducted l to the loop seal water. However, due to field installation constraints, the loop seal piping temperatures were not as high as expected. The measared temperature profiles at the three loop seals for the Units 1 and 2 PSARV systems fall between the bounds of .
bot and cold.
The actual measured loop seal temperatures are tabulated in Table 1. De node notations are based on the Figure 1 convention. It can be seen that the temperatures WPF127M/052192:10 3
- . . . . . - - -- -~. - - -- . .~ .- .- - - _.- . - - - . ..
. Westinghouse Proprietary Clau 3
, are higher upstreara of the loop seal which Is closer to the pressurizer and the l temperatutes are colder near the inlet of the safety valves.
Variable Loop Seal Temocraturea ,
Based on the temperature data in Table 1 the fluid properties (enthalples) of the condensed liquid water are obtained. Two cases are considered one case with the ,
average fluid properties in each loop seal and the other case with vailable fluid i properties along the loop seal piping in each loop seal. Thermal hydraulle analysis are performed for these two cases. The thermal hydraulic analysis model used is the same as the 1982 thermal hydraulic analysis model and is shown in Figure 2.
Comparison of Thermal Hydraulic Forces The thermal hydraulle forces ger.erated by the above two cases are compared with the
- forces generated from the 1982 hot and cold loop seal thermal hydranile analyses. The -
forces are tabulated in Table 2. Since the forces for the variable loop seal temperature are more conservative than those for the average loop seal temperature and the -
condition reflects the actual fiehl condition, they are used to perform the time history structuras e aalysis of the pressurizer safety valve piping system.-
Thermal Hydraulic Analysis Computer Programa The computer program used for the thermal hydraulic analysis is ITCHVENW'l, cy=2r
' 'Ihis program was upgraded several times since 1982 and was renamed to ITCHVENT from program ITCHVALVE '03.- l Program ITCHVALVE was benchmaked against the EPRI test data. -ITCHVALVE is a-1.D thermal hydraulic' code that calculates the time history fluid properties within the PSARV system for the condition when the safety or relief valves open. The thermal hydraulic forces e.re calculated by anot> r program- ,
called FORFUNIl, con'idering the momentum changes for the fluid in each element of l
WPF1276J/G52t92:10 4
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Westinghouse Proprietary Class 3 the piping segment. A typical FOF FUN output of thermal hydraulic time history force at force nurnber 14 for Faricy units is illustr:.ted in Figure 3.
2.2 Structural Modeling The structural modeling and analysis of the pressurizer safety valve piping system were performed using the WECAN Computer Codel"I. The piping system was modeled by p.pe, elbow, support stiffness elements with both elastic and clastic / plastic capabilit!es.
Consistent mass effect was considered in the analysis. For the analysis of the piping system witn combination of deadweight and safety valve thrust discharge loadings, WECAN dynamic transient time history analysi!. option was chosen. The input time.
history was determined by ITCHVENT and FORFUN computer programs and was applied to the piping system structural model.
Figure 3 shows the structural model of the Unit 2 safety line system, which contains three 6 inch safe valves on three lines before meeting a 12 inch common header. The 12-inch common header leading to the pressurizer relief tank is also in the model. Part of the relief line piping was modeled in the structural system to account for the structural system interactions. It should be no'ed that due to the similarity in layout between two units, only Unit 2 was modelled and analyzed. Ratioing technique was developed and used for the Unit I structural evaluation.
2.3 Elastic Pi.ntic Analysis Method With respect to the clastic plastic : .1alysis, the WECAN procedures are based on the j incremental theory of plasticity using Von Mises yield condition and the associated flow l rule, the Prandtl Reuss equations. In the analysis the combined isotropic and kinematic hardening rule was selected in' conjunction with the multilinear stress-strain curve (see Figure.5). The initial strain method or the method of successive clastic approach were adopted so that the stress strain relation stays the same throughout the computation.
WPFt2763/052292:10 5.
Westingbouse Proprietary Class 3 Therefore, the effect of plastic deformation during a load incremen. is taken into account by introducing a set of fictitious body forces into the equations.
The time history solution for the dynamic thrust analysis of safety valve discharge with loop seal water slug was obtained from WECAN computer programs using direct integration methods. Since the purpose of this analysis is to determine the inelastic behavior of the piping system under the extreme loading of valve thrust, the plastic option of the WECAN program was used. The resulting stress and strain at 8 equally _
spaced circumferential points of a given ci Ms section were calculated for a 0.5 second time history following the safety valve discharge action.
Since the thrust loading is applied very rapidly and the piping material behaves inelastically, the integration step size was an important factor to achieve a converged and -
realistic solution. Solution accuracy can be affected by the selection of the number of load steps in time, which defines integration step size, at, and by the number of iterations within each load step. The number of load steps controls the rate of loading from time history, while the iteration number controls the convergence in non linearity between load steps. For this analysis,it was determined that AT = 0.33 m-second was sufficient to provide a converged solution.
3.0 PIPING COMPONENT AND SUPPORT EVALUATION 1
3.1 Piping Component / Support Systems Evaluation Criteria The pressurizer safety and relief valve piping system was originally qualified to its design basis r.llowables prior to 1980 TMI requirements. The design basis was the requirements
( of ASME B&PV Code Section III,1971 edition, including summer 1971 addenda for l Class 1 piping and the ANS B31.11967 Code with 1971 addenda for the NNS piping. In 1982, Westinghouse performed additional evaluation to address TMI related issues by considering the cold loop seal loads for these pipine, systemslR Criteria used in that analysis was based on the recommendation from piping subcommittee of tne PWR WPF1276J/052292:10 6
. ... , , e, Westinghouse Proprietary Clus 3 PSARV test program and was documented in a WCAP.101051"l. That criteria was reviewed and accepted by the NRC in a 1986 SER14 In this evaluation, the same loading combination and piping evaluation :riteria as the ones in WCAP 10105 were used. This load combination and evaluation criteria are provided in tables 3 and 4.
The piping / support system was clastically qualified for all loadings without considering the valve thrust loads. For the emergency condition in Tables 3 and 4, where safety valve thrust loads are required to be combined with other normal leads, simultaneous safety valve discharge can result in a potential overstress in the piping / support system at discrete locations downstream of the safety valves. Therefore, the elastic plastic analysis- ,
technique was used to determine if sufficient reserve margin under plastic action such that the pipe would not result in structural failure. The following structural integrity evaluation criteria can be used in this analysis for both Class 1 and NNS portions of the piping system if der.med necessary:
i e Piping components strain limits from 1986 ASME Code Case N e Pipe support - structural integrity limit 3.2 Piping Component Evaluation Results 3.2.1 Piping Components Using clastic analysis tnchniques, the Class 1 piping _(which connects the pressurizer
~
safety line nozzle to the 6' safety valve), were qualified to the allowables listed inTable 3 with the effect of valve thrust under emergency condition. However, using clastic analysis techniques, small portion of the NNS piping, (which connects the 6" safety valve and the relief tank) cannot meet the allowables listed in Table 4 for the same effect of valve load thrust under emergency conditions (see Table 6).
WPF12763/052292:10 7
.- - . . - - =- .
Westinghouse Proprietary CI:ss 3 Under emergency load combination, elastically calculated over stresses were predicted at certain piping components (such as 6" x 12" branch connections and 6" x 12" reducer along the common header). , wever, the results from clastic plastic piping system analysis showed a rnaximum total strain of 0.5% at the reducer; 0.3% at branch connection and 0.2% at some elbows. All of these strains are very small compared to the Code Case N-47 limits which permit a maximum of 1% for membrane,2% for membrane plus bending, and $% for local discontinuity. More importantly, the above calculated total strain values occurred only at outside surface point along the circumference of the pipe cross section. All other paints along the circumference have much smaller value in total strain, herefore, under such small strain, the discharge piping downstream of the safety valve will not rupture and the structural integrity of the pipe will be maintained. Observation from the clastic plastic analysis results also indicated that the cross section of the pipe was not severely deformed and will not affect the safety valve operation, since the pipe cross section area will not become restricted, 3.2.2 Safety Valve Nonles One additional means to ensure that the safety valve remains operable after the loop seal water is discharged is to assess the valve noule loads with respect to the valve
! operability limit provided in the equipment specificationt'31. For emergency coadition, the calculated valve noule loads from the combination of deadweight, pressure and valve thrust effects are within the equipment specification allowable. This allowable requires the maximum total valve noule stress to be 75% of the yield stress of the nonle material at temperature. In addition,it further requires that the maximum bending stress be 50% and the maximum torsion stress also be 50% of the yield stress of the nonle at .emperaturc.
1 WPF1276!/052292:10 8
Westirighouse Proprietary Class 3 l
3.3 Support Comoonent Evaluation Results 3.3.1 Loading and Load Combinations The piping system loading conditions considered for the pipe support evaluation consisted of the valve thrust loadings discussed above in combination with tne existing design basis dead weight, normal thermal expansion, transient thermal expansion, and the OBE & SSE seismic event loedings.
Since the pipe supports had previously been qualified for the Normal, Upset, Emergency, and Faulted conditions, the supports were only evaluated for the worst case load combination including the valve thrust loads from the clastic plastic piping system analytis. The loading combination used for support evaluation is:
P=DWtnm m 1 lSSE' + nrst' 3.3.2 Stress Acceptance Criteria .
The purpose of the support evaluation was to demonstrate that the suppons retained their integrity for the controlling combined loads as discussed in Section 3.3.1. This was accomplished by generally limiting the actual support member stresses to the allowable I stress limits established by the ASME Boiler and Pressure Vessel Code,Section III, i l Subsection NF and Append F,1974 Edition. The code of record, AISC 7th Ed., does not address the faulted loading combir,ation. ASME Subsec. NF was then used for the 1992 evaluation since it is nearly the same as AISC for the normal and upset conditions, t
and it provides for the extreme faulted loading combination. In addition, the Subsec. NF criteria is consistent with the support criteria of most other nuclear plants. For a few standard component support parts, actual stresses were compared with the ultimate strength of the support meterials. It is noted that the support stress limits used insure overall clastic behavior of the supports, even though maxhnum stresses in some outer l
fibers'of the support elements may exceed the material minimum specified yield strength of the material.
WPF127M/052192:10 9 j
Westinghouse Proprietary Class 3 Concrete expansion anchors were limited to manufacturer's allowables with a Factor of Safety of 1.40 in some cases in lieu of 4.0.
3.3.3 Results
, Class I supports -- the results of the pipe support evaluations based on the as built suppo.'t data provided to Westinghouse show that all the Unit I and Unit 2 pipe support standard Grinnell components.I"I structural members, and base plate element "I1863 l stress levels are within the allowable stress limits of ASME Subsection NF and Appendix F and will maintain their structural integrity and stability for the specified loading combination.
NNS supports - nearly all the Unit 1 and Unit 2 supports satisfied the ASME .
Subsection NF and Appendix F stress criteria. Stresses in a few Grinnell standard components (sway struts and a snubber rear bracket) exceeded the NF/ App. F criteria, but were well within the minimum specified ultimate strength of the material. All the NNS supports will maintain the;. structural integrity for the specified loading combination.
4.0
SUMMARY
AND CONCLUSIONS ne purpose of the analysis and evaluation described in this report is to address the concerns identified in the NRC Safety Evaluation Report (SER) published in 1986 concerning NUREG-0737, item II.D.l'I. l As discussed in the introduction of this report, subsequent to the NRC SER, a variable loop seal condition in the pressurizer safety lines has been achieved due to the modification of insulation on the pipe. As a result of the new insulation, water temperatures in the loop seal were increased. Considering the increased loop seal temperatures new thermal hydraulic loads were generated. The -
variable loop seal reflected a more realistic yet conservative condition in comparison with EPRI test results.
WPFt276J/052292:t0 10
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Westinghouse Proprietary Class 3 With these time history thermal hydraulic thrust loads applied simultaneously to the three safety line system, the clastic responses of the system meet all the stress allowables listed in Table 3 for Ciass 1 piping. However, the same clastic responses showed some over stress (compared to the Table 4 allowables) in NNS portions of piping components located downstream of safety valves, for the load combinations of emergency condition.
Following the clastic sy tem analysis, an elastic-plastic piping system time history analysis -
was performed using WECAN computer code. The results from this analysis demonstrated that the over stress from elastic responses produced very little plasticity.
The maximum total strain at the worst location in the whole system was only 0 5% of strain, which occurred at a single point on the outside surface of the pipe cross section.
The next highest total strain locations were at a branch connection with a value of 0.3%
and at some elbows with a value of 0.2%. All of these strains are very small compared to the ASME Code Case N 47 limits which permit a maximum total strain of 1% for membrane,2% for membrane plus bending and 5% for local discontinuity. More importantly, the abovc. calculated total strain occurred only at outside surface point along the circumference of the pipe cross section. The remaining points in the cross section have very little strain such that no cross sectional area change was possible. Hence, under the emergency condition analyzed, it is demonstrated that the structural integrity of the pipe was maintained. Furthermore, the cross sectional area remained the same, so that the operability of the safety valve under the thrust loads is assured.
In addition to the pipe, all pipe supports and their structural embedments were evaluated to their structuralIntegrity limits and found acceptable.
5.0 REFERENCES
- 1. NUREG 0737," Clarification of TMI Action Plan Requirements," NRC,
, November 1980.
WPFt276J/052292:10 11
Westinghouse Proprietary Clau 3
5.0 REFERENCES
- 2. " Application of RECARS/ MODI for Calcula'. ion of Safety and Relief Valve Discharge Piping Hydrodynamic Loads," EPRI NP 2479, Final Report, Dec.1982.
- 3. Letter, F. L Clayton, Jr., to S. A. Varga, NRC, " Response to NUREG.
0737/NUREG 0660 TMI Action Plan Requirements," April 1,1982.
- 4. Letter, F. L Clayton, Jr., to S. A. Varga, NRC, " Joseph M. Farley Nuclear Plant -
Units 1 and 2 NUREG-0737, item II.D.1 Response," July 1,1982.
- 5. 12tter F. L Clayton, Jr., to S. A. Varga, NRC, " Joseph M. Farley Nuclear Plant, Units 1 and 2 NUREG-0737, item II.D.1," November 4,1982.
- 6. I etter, E. A., Reeves, NRC to R. P. Mcdonald, Alabama Power Co.," Completion of Review cf item II.D.1 NUREG-0737 Safety and Relief Valve Testing for Joseph M. Farley Nuclear Plant Unit Nos.1 and 2," December 16,1986.
- 7. Letter, R. P. Mcdonald to L S. Rubenstein, NRC, " Joseph M. Farley Nuclear Plant Units 1 and 2 Completion of NUREG Item II.D.1 Review," February 5, l
1987.
- 8. Letter, W. G. Hairston, III to NRC, "Josed M. Farley Nuclear Plant, Units 1 and 2 NUREG 0737. Item II.D.1, Review Conipletion Schedule Uphte," September 16,1988.
i l
l
- 9. Letter, J. E. Garlingtor, to J. A. Knochel, Westinghouse, Activity Code 52073, ES-912073, " Pressurizer I/r ; ic"1 Analysis," Navember 27,1991.
WPF12767/032192;10 12
,s :) m e, Westinghouse Proprietary Class 3
- 10. (a) W Report WCAP 9924,"ITCHVALVE Code Description and Verification", M. A. Berger & K. S. Howe, July,1982.
(b) Westinghouse Internal Letter SE&fT CSE-4651/31/90," Release of Thermal Hydraulic and Related Computer Codes (lTCHVALVE, ITCHVENT, FORFUN, KJTRPLT2) for Production Use on the Cray X.
M P."
- 11. Ni Computer Program WECAN, WECAN/PLUS User's Manual, Dec.1,1990, First Edition, Westinghouse Electric Corp. Pittsburgh, PA.
- 12. Westinghouse Report WCAP-10105," Review of Pressurizer Safety Valve Performance as Observed in the EPRI Safety Valve and Relief Valve Test Program", June,1982.
- 13. Westirghouse Equipment Specification #952445, Rev.1. March 31,1977.
- 14. Letter from Frank Birch, Grinnell Corp. to J. Himler, Westinghouse,
- FB1V/1940D,"Farley Units 1 and 2 Project," April 3,1992. '
WPF1276J/052t92:10 13 e
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Wettinghouse Proprietary Class 3 TABLE 1 .
4 LOOP SEAL TEMPERATURE DATA ;'
(DEG. F) t Farlev Unit 1 Farley Unit 2 T
LQC LQQP_& LOOP B LOOPI LD.C LOOPA LOOP B LQDf_C !
1 653 653 653 1 653 653 653 l
2 560 552 536 2 548 529 '557 3 502 470- -432 3 483 440 495~
4 457 424 350 4 465 450 467- ,
~$ .457 424 350- '5' 465 - 450 467 G 404 343 289 -6 4W 410 397 .
7 404. -343 289 7 '416 410- 397 -
4 8 283 228 210' 8 290 -290'- 276
[
9 160 144 141 9 151 148 154 h
~
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e
.f f
b 4
L WPF1276J/052192:10 14 -
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Westing, house Proprietary Class 3 TABLE 2 FORCE COMPARISON FOR DIFFERENT LOOP SEAL CASES FORCE liGI AVERAGE VARIABLE COLD 1 156 452 392 157 2 193 544 492 194 3 170 485 439 171 4 4774 7970 7057 4798 5 4S69 9251 7687 4877 6 4941 % 33 8129 4944 7 2945 8647 8125 4960 8 1714 3403 4279 7162 9 13747 13374 27888 37205 10 6880 8272 9608 36778 11 25330 29502 35571 65200 12 10947 18510 18223 24279 13 17114 19192 20568 22811 14 38405 47330 51739 51570 15 12907 13648 14399 8846 16 155 453 374 157 17 192 546 470 194 18 169 486 420 171 19 4769 7818 6877 4799 20 4869 9057 7327 4872 21 4946 9424 7636 4948 22 3395 11586 10853 6528 23 13397 13958 31174 37229 24 10339 8935 10672 39906 25 155 452 370 157 26 192 544 465 194 27 169 485 416 171 28 4 776 7622 6719 4803 29 4878 8835 7tMS 48E1 30 4958 9164 7277 4%3 31 3240 11772 10083 5956 l
32 10776 15255 29570 29330 33 9127 4463 9976 35410 34 4904 3298 5765 38350 35 4702 3200 $366 39279 36 9522 7204 11023- 35707 i
WPF1276J/052192:10 15
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.,..:g Westirghouse Proprictary Class 3 TABLE 3 LOAD COMBINATIONS AND ACCEPTANCE CRITERIA FOR PRESSURIZER AND RELIEF VALVE PIPING UPSTREAM OF VALVES CLASS 1 PIPING Piping Plant / System Allowable Stress Operatine Condition Load Combinalian Intensity Ql.. .'
"Jy No" N ?.5 S.
d D, N + OBE 1.5 S.
-: Upse' N + SOTu 1.5 S.
Upset N + OBE + SOTu 1.8 S,/1.5 S/3)
Emergency N + SOTe 2.25 S /1.8 Sg3)
M Faulted N + SSE + SOTr 3.0 S.
1-uS- (1) See Table 5 for definitions of load abbreviations
, (2) Use SRSS for combining dynamic load responses.
- (3) The smaller of the given allowable is to be used.
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WPF1276J/052192:10 16
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. ! . '. a t - e *, .I Westinghouse Proarietary Class 3 ' -j
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- ' TABLE 4; -j
- i. LOAD' COMBINATIONS AND ACCEPTANCE CRITERIA FOR ~
h
- l. PRESSURIZER SAFETY AND RELIEF VALVE PIPING -
i-DOWNSTREAM OF VALVES :-1 i NNS PIPING ~ j 4
- i
$ 4 1
- Plant / System- Piping Ooeratine Condition 1,oad Combination Allowable Stress
[ Normal Upset N'
11 + OBE '
1.0 S --
1.2 S.
i~ Up;et N + SOTu . 1.2 S L Upset N + OBE !+ SOTu - -1.8 S,;
Emergency N + SOTS 1.8 S,
~
- . Faulted - N + SSE + ' SOT, . 2.4_ Sn -
l NOTE?: - (1) See Table 5 for definitions of load abbreviations-i- -(2) Use SRSS for combining dyncmic ' bad responses.-
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l -
F
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- 10 - 17
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TABLE 5-DEFINITIONS OF LOAD ABBREVIATIONS N =. Sustained loads during normal plant operatinrt-SOT- = System operating transient- '
SOTu = Relief valve discharge transient 3
SOTS = - Safety valve Jischarge transient .
SOTr = Max (SOTu; SOTch or tramition Dow OBE = . Operating basis earthquake SSE = Safe shutdown earthquake- ,
= Basic material allowable stress at maximum (hot) temperature .
S. .
S, = Allowable design stress in'ensity S, = - Yield strength value .
I t
r L
WPF1276J/052i92:10 18 l-3
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We'stinghouse Proprietary Clau 3 TABLE 6 -
FARLEY UNITS 1 AND 2 SAFETY LINE PIPE STRESS AND STRAIN
SUMMARY
FOR EMERGENCY CONDITION.
Code- Code Node Piping Maximum - - Allowabla. -Total -
P.Dini Comoonents Stress (ksi) * -- Stress (ksi) Strain (c'i+ c 1290' Butt weld 18.15 36.225 ~0 1460' Long radius ;31.65 36.225 0.15 %
elbow 100 " Branch 49.25 33.84- 0.26 %
connection
' 690 "
Reducer 71.08 L -33.84 0.46 % -
1490 " Welded 54.97 55.42+ - -
. attachment
@ support R120
- ASME Class 1 piping, upstream of safety valves
" ASME NNS piping, downstream of. safety i dves
- From elastic analysis
+ From elastic plastic analysis
+ Based on ASME Code Case N-318 allowable
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