ML20235J375
| ML20235J375 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 08/31/1987 |
| From: | ABB IMPELL CORP. (FORMERLY IMPELL CORP.) |
| To: | |
| Shared Package | |
| ML20235J333 | List: |
| References | |
| 09-0870-0014, 09-0870-0014-R01, 9-870-14, 9-870-14-R1, NUDOCS 8710010416 | |
| Download: ML20235J375 (117) | |
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1 EVALUATION OF PRESSURIZER SAFETY AND RELIEF VALVE SYSTEM FINAL REPORT Submitted to Hisconsin Electric Power Company Prepared by Impell Corporation 300 Tri State International Suite 400 Lincolnshire, IL 600lS Impell Report No. 09-0870-0014 Revision 1 August 1987 e710010416 070925 . - -
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REPORT APPROVAL COVIR SHIET lliant: _ Wisconsin Electric Power Company Yoj;ct SRV Evaluation Job .%tz ber: 0870-005. 006 Evaluation of Pressurizer Safety and Relief Valve System Final Report for
$ rt 2218: Point Beach Nuclear Plant leport Nu:cher: 09-0870-0014 Rev. O _
20 work desce. bed in its Report was performed in accords:ce wtth the Impe11 Ns.lity Assurs:ce Progra.:n. De signatares below ver:fy te accur:cy cf dis Report ad its compliance with applicable quality assurance We=ents.
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Impell Report No. 09-0870-0014 Revision 1-TABLE OF CONTENTS P191 LIST OF TABLES i LIST OF FIGURES 11
1.0 INTRODUCTION
1 1.1 General 1 1.2 Hork Performed 1 1.3 Modification of Loop Seal 2 1.4 Conclusions -3 2.0 SYSTEM DESCRIPTION 4 t
2.1 General 4 2.2 Piping System 5 2.3 Operating Conditions 5 3.0 POINT BEACH NUCLEAR PLANT EVALUATION 7 31 Introduction 7 3.2 Safety Valve Evaluation 7 3.3 Thermal-Hydraulics Analysis 14 3.4 Piping Evaluation 19 4.0 POINT BEACH NUCLEAR PLANT MODIFICATIONS 25 '
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4.1 Introduction 25 4.2 Methodology 25 4.3 Piping Modifications 25 5.0 RESULTS 27 )
5.1 Piping Stresses 27 5.2 Nozzle and Valve Flange Load 27 5.3 Valve Accelerations 27 5.4 -Support Loads 27
. 5.5 Discussion of Piping Results 27 REFERENCES 30 APPENDIX A: Description of Computer Programs APPENDIX B: SUPERPIPE Models i
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'Impell Report No. 09-0870-0014 Revision 1 TABLE OF CONTENTS (cont'd)
APPENDIX C: Detailed Pipe Stress and Support Load Summaries-APPENDIX D: Insulated Box Drawings s
APPENDIX E: Support Drawings f
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o Impell Report No. 09-0870-0014 Revision 1 LIST OF TABLES EME2Sf7?ILMC2EIE2722E8dE22Isil2EE12[5I3133 Iahle Title 2-1 . Safety Valve Parameters 2-2 Power-Operated Relief.and Block Valve Parameters 3-1 Applicable EPRI Tests for PBNP Safety Valves 3-2 Comparison of Results for Applicable EPRI Tests 3-3 Maximum Calculated Backpressure
'3-4 Maximum Calculated Temperatures 3-5 Load Combinations for Piping Analysis 3-6 Pipe Support Load Combinations 3-7 . Allowable Stresses for Seismic Class Piping 3-8 Allowable Stresses for Non-Seismic Class Piping 4-1 Unit 1 Support Configuration 4-2 Unit 2 Support Configuration 5-1 Unit 1 Maximum Pipe Stresses 5-2 Unit 2 Maximum. Pipe Stresses 5-3 Nozzle / Flange Loads 5-4 Safety Valve Accelerations 5-5 Unit 1 Piping Support Loads 5-6 Unit 2 Piping Support Loads j g, w ,, , , y. . ,, ,.
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Impell Report No. 09-0870-0014 Revision 1 LIST OF FIGURES i
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1 Fiaure Title 2-1 PBNP Unit 1 Piping Configuration 3-1 Loop Seal Box Insulation 3-2 Loop Seal Temperature Profile 3-3 Typical Plot of Safety Valve Backpressure 3-4 PBNP Unit 1 REFORC Model 3-5 Unit 1 Force Time History for Data Point F-1 3-6 Unit 1 Force Time History for Data Point F-17 3-7 Unit 1 Force Time History for Data Point F-5 3-8 Seismic Response Spectrum il f'- ^
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Impell Report No. 09-0870-0014 Revision 1
1.0 INTRODUCTION
m..am.s.amawammaa.m.muuasam u maaw w m e s m m y n men w nwgnema 1.1, General Impell has completed an evaluation and required modification of the Point Beach Nuclear Plant (PBNP) pressurizer safety and relief valve system. This evaluation was performed for Hisconsin Electric Power Company in accordance with the recommendations of NUREG-0578, Section 2.1.2, clarified by NUREG-0737, Item II.D.1, and by the NRC's letter of September 29, 1981.
This report summarizes the evaluation and modi fication.
1.2 Work Performed The initial Impell scope of work included an evaluation of the functionality of the safety valves and the functionality and integrity of the system piping. The scope of work was later expanded to .'
include the design of pipeline modifications required to meet allowable design limits. The operability of power-operated relief valves (PORV's) and the block valves was not evaluated within this scope of work.
PORV and block valves operability has been addressed in Wisconsin Electric Power Company letters to the NRC dated June 30, 1982 and August 9, 1982.
The operability of the safety valves was evaluated principally by correlation with the industry-sponsored Electric Power Research Insitutee (EPRI)
Test Program.8 The applicability of this program to PBNP is addressed in Section 3.2.1. The results of the evaluation are given in Section 3.2.2.
For the pressurizer safety and relief valve discharge piping system, therenal-hydraulic analyses were performed to calculate the bounding dynamic loading induced on the piping by rapid valve actuation. The computer program RELAP5/H0012 w(s used, together with the post-processor REFORC.3 These analyses ,
are described in Section 3.3.
Piping analyses were performed for these thermal-hydraulic loads, using the computer program SUPERPIPE4 Analyses were also performed for gravity, thermal, pressure, and seismic loads. These w . gv - n, , , ,
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Impell Report No. 09-0870-0014 Revision 1
1.0 INTRODUCTION
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" uduGin MhLo,Gutuus l shnaku,Kan nuakaan,u analyses are described in Section 3.4. The results from these analyses were used as a basis for the design of the modified support scheme.
The modified support scheme was determined by iterating on possible support locations to establish the optimal configuration. A full walkdown of the piping system was performed by Impell engineers to gather information on existing support structures and facilitate the location of new supports. Resulting support loads for the modified support scheme were used to qualify existing supports (when possible) and to design new or modified supports. Detailed calculations and drawings were prepared for the construction of new and modified supports.
During recent outages, Impell provided on-going construction support to expedite the installation effort. Hinor construction alterations were evaluated and incorporated into as-built calculations and drawings.
1.3 ' Modification of As an ongoing part of the evaluation, potential of Loon Seal modifications to the loop seal and the safety valves were considered as a means of either improving system performance or reducing discharge piping loads.
Based on the adequacy of the original system, no modifications were made to the valves. Furthermore, the loop seals have been retained. However, Hisconsin Electric has elected to raise the temperature of the loop seal water by adding insulation upstream of the safety valves.
Calculations were performed and drawings prepared for construction of the insulated boxes used to insulate the lo,p seals. Hinor construction alterations of the insulated boxes were incorporated into as-built calculations and drawings.
The evaluation of safety valve operability (Section 3.2) is based on the cold, uninsulated loop seal. '
Raising the loop seal water temperature will further assure the valves' operability.
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Impell. Report No. 09-0870-0014 Revision 1
1.0 INTRODUCTION
i ga> ~gww nm.yma m ?~ggympfwwmmw m, saww=ax..aaukassanumm&myrcp;w >yw wmA3 As the decision to insulate the loop seal was made before the thermal-hydraulic analyses were started, these analyses (Section 3.3) have been based on the modified loop seal temperature profile. Thus, they were used as the basis for the modification to the discharge piping system.
The discharge piping stresses reported in Section 5.0 for the modified system are based on the modified l
temperature profile.
1.4 Conclushni Based on the EPRI Test Program results and the plant-specific evaluation described in Section 3.2, it is concluded that the operability of the PBNP safety valves is confirmed. For the postulated, severe system operating transients under which they may be activated, the safety valves should relieve pressure and prevent overpressurization.
Furthermore, their operating characteristics are such that the conclusions drawn with regard to the safety aspects of PBNP in Section 14 of the FFDSARS should not be impacted.
Due to the high out-of-balance loads induced by the sudden discharge of the water loop seal, the stresses calculated on certain portions of the original discharge piping system exceeded the specified allowable stresses. The modified support scheme Dista11ed on the discharge piping system reduces pipe stress to within allowable limits. Support loads generated from the modified support scheme are now within the capacity of the respective supports.
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2.0 SYSTEM DESCRIPTION
[EEEEE E MEC 12EMKIE5F M 2[$1 1538 2.1 General PBNP is a two-unit power plant. Each unit is a I Westinghouse pressurized water reactor with two primary coolant loops. Each unit is rated at 1518 MHt . In general, Unit 2 is a mirror image of Unit 3 1.
l The safety and relief valve system for each unit I consists of:
Two spring-loaded Crosby Valve and Gage Co.
HB-BP-86 series safety valves
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Two Copes-Vulcan Inc. power-operated relief valves j Two Velan gate-type block valves with Limitorque operators Three pressurizer outlet piping lines: one
, line, including a loop seal, for each of the two safety valves, and one (branching) line for the block and power-operated relief valves A common discharge piping line, terminating in the pressurizer relief tank (PRT)
Details of the safety and power-operated relief valves are included in Tables 2-1 and 2-2 respectively.
The inlet and outlet piping is described in Section 2.2.
The layout of the Unit I system is shown in Figure 2-1. The Unit 2 layout is essentially a mirror image of Unit 1, with some difference in the piping near the PRT.
If an abnormal transient causes a substained pressure increase in the pressurizer at a rate exceeding the control capacity of its spray system, a high-pressure trip signal is activated. This signal opens the PORV's. If the pressure continues to rise and g .
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2.2 Pioina Svsiga Each safety valve is connected to a pressurizer outlet nozzle by 4-inch diameter piping in a loop seal configuration. The loop seal was originally uninsulated and contained cold water. For the modified system, the loop seal is insulated and contains hot water. Each safety valve also has a 6-inch discharge (or tail) pipe which runs into a common 8-inch header pipe.
The two PORV's share one pressurizer outlet nozzle.
A 4-inch pipe from this nozzle branches into two i 3-inch pipes, one for each PROV. For Unit 1, one l PORV is a 3-inch, the other a 2-inch valve. For Unit 2, both are 2-inch valves. The 2-inch PORV's are attached to the 3-inch pipes through 3 by 2-inch reducers. Each PORV has a 3-inch discharge (or tail) pipe. These run into a common 4-inch header which runs into the 8-inch header common to the safety valve discharge piping.
The block valves are in series with (and upstream of) the PORV's.
The 8-inch header pipe discharges into the PRT. The PRT has a volume of 5984 gal and is equipped with an L-quencher and a 100 psig rupture disc.
1 2.3 Qoerating Conditions The most severe reactor coolant system overpressure condition requiring operation of the PORV's or of the ,
PORV's and the safety valves would occur following (
the postulated instantaneous seizure of a reactor coolant pump rotor - a " locked rotor" accident.5,6 {
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The transient analysis for a locked rotor accident l' conservatively assumes that the PORV's do not operate and that pressure relief is through the safety valves only. The peak pressurizer pressure for this case is 2763 psig with a maximum pressure ramp rate of 297 psi /sec.
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ObdhAih5Lawas.5:4MSM mn numm Upon clearing of the loop seal water, the fluid condition is saturated steam only. No postulated PBNP transient results in the passage of solid water through.the safety valves after discharge of the loop seal water. Cold overpressurization may result in passage of solid water through the PORV's only; however, given the slow opening' time of the PORV's relative to the safety valves, the locked rotor transient would induce a much more severe loading of the system piping.
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Impell Report No. 09-0870-0014 Revision 1 3.0 POINT BEACH NUCLEAR PLANT EVALUATION
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- n. , ~:a n aw. w w ~ w.ax wa.wa 3.1 Introduction The evaluation was performed in three parts.
First, the operability of the PBNP safety valves was evaluated by correlating the EPRI Test Program data
- to the PBNP-specific design.
1 Secondly, thermal-hydraulics analyses were performed to determine the bounding forces imposed on the piping by valve actuation. Actuation of a valve allows the discharge of loop seal water and high-pressure steam from the pressurizer into the discharge piping, inducing pressure and momentum transients. Until steady-state is achieved, these transients create significant unbalanced forces on each straight run of the piping.
I Thirdly, dynamic piping analyses were performed to determine the response of the piping to these (and other relevant) loads. From these analyses, upper-bound stresses on the piping and upper-bound loads on the supports were calculated.
These analyses are described below.
3.2 Safety Valve Evalu- At the request of the PHR utility industry, EPRI ation directed a full-scale test program to evaluate the performance of pressurizer safety and relief valve piping systems. A number of valves, representative of those currently installed, were tested under conditions that encompassed typical, postulated pressure relief transients. Actual testing was completed in December 1981. Reports on the results of the tests have been issued to the participated utilities and the NRC. 7,8,9 3.2.1 Applicability of The EPRI test program included tests on a number of LPRI Test Resulti Crosby and Dresser spring-loaded safety valves.
These valves were tested with various inlet piping configurations and for various fluid and flow conditions. Those tests which are relevant to PBNP h
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and their applicability are discussed in this section. Valve type and installation, inlet and j outlet piping configuration, and operating conditions ;
are addressed.
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Crosby 3K6 and 6M6 safety valves were tested. These i valves are structurally and functionally similar to {
the PBNP Crosby 4K26 valves. A comparison was !
performed by Crosby Valve and Gage company and is '
included in the EPRI Valve Selection / Justification Report.10 This report considered the effect on valve operability of differences in valve operational characteristics, materials, design details, and size. The conclusion states that the selected test valves (3K6 and 6M6) do represent (and thus the EPRI test results are fully applicable to) the Crosby valves presently installed in PHR plants (including the 4K26).
Inlet Pioina Configuration The Crosby valves were tested with both short and long (loop seal) inlet piping configurations. Both PBNP units have long inlet, loop seal configurations.
The geometry of the long inlet configurations used for both the 6M6 and the 3K6 tests was essentially the same as that installed at PBNP - however, specific pipe dimensions and lengths differed. For example:
The 6M6 and 3K6 inlet piping was of 6 and 3-inch diameter, respectively: The PBNP piping is 4-inch diameter.
The distance from the pressure source nozzle to the valve inlet was approximately 142 and 114 inches for the 6M6 and 3K6 valves, respectively: For the PBNP units, it is less than 100 inches.
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La Aa.=win.;ne A n.c rm n dmeymumm a n s 64 The volume of the loop seal water was approximately 1760 and 470 cubic inches for the 6M6 and 3K6 valves, respectively: For the PBNP units, it is less than 370 cubic inches. l 1
Similarly, the lengths of the loop seal water slug were 94 and 61 inches for the test valves, i but less than 48 inches for the PBNP units. ]
In summary, while the 6M6 and 3K6 tests are clearly applicable, it is noted that the PBNP water loop seal slugs are much smaller than those used in the Test !
Program. Also, the inlet piping length is shorter - '
this will lead to the plant-specific inlet pressure drop being generally less than the test inlet pressure drop.
Outlet Pioina Configuration The test configuration is essentially similar to that for PBNP. However, the distance from the valve outlet to the first elbow was considerably more than that for PBNP (approximately 59 inches versus 22 inches). Similarly the second straight run was longer (approximately 280 inches versus 60 to 110 inches). The effect on valve operability is covered under the discussion of backpressure. However, the shorter runs at PBNP will result in generally lower transient out-of-balance forces on the outlet piping than were recorded in the tests.
Fluid Conditions 1 The valve inlet fluid conditions used in the loop seal tests included cold loop seal water followed by saturated steam, saturated water, or subcooled water. Transitions from steam to water, and heated loop seal water cases were also run.
For the current PBNP configuration, the tests with a cold loop seal followed by saturated steam only a'e r applicable - for the modified configuration (see Section 1.3), the tests with a heated loop seal p - .<
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water followed by saturated steam are also applicable. These tests are listed on Table 3-1.
IDiet Pressure for the PBNP safety valves, the analysis of the critical, locked rotor transient (given in Reference .
- 5) is based upon the safety valve opening at 2485 /
psig. The calculated pressure ramp rate at this 'p pressure is 297 psi /sec., and the peak pressurizer pressure is 2763 psig. These are the maximum ramp rate and peak pressures for any transient analyzed.
All the tests listed on Crosby safety valves were on valves set at 2485 psig. The pressure. ramp rates range up to 375 psi /sec. Clearly, these inlet pressure test conditions envelope those under which the PBNP safety valves may be required to operate.
Backpressure The steady-state backpressure for the EPRI tests )
listed in Table 3-1 ranged from 227 to 700 psia. The )
thermal-hydraulic analyses described later in this report determined backpressure for the PBNP valves of approximately 550 psign Aeolicable Testt Table 3-2 includes a comparison of pressure ramp rates and backpressure for the relevant tests listed in Table 3-1. Based on these, tests which are directly relevaat to the PBNP configuration (specifica11y, those with high pressure ramp rates and backpressure) are indicated. While the remaining tests in this table do provide applicable i data on the operability of the PBNP valves, they are of less direct relevance.
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Impell. Report No; 09-0870-0014 Revision 1 3.0 POINT BEACH NUCLEAR PLANT EVALUATION E3222Ei?EEEC0CO2211TC%If552ME3I31Il 3.2.2 PBNP Safety Valve The Crosby 6M6 and 3K6 safety valves operated Operability successfully in all tests applicable to the PBNP loop seal configuration - a water loop seal followed by saturated steam. In all cases they relieved pressure and prevented excessive overpressurization.
The applicability of these tests to PBNP has been addressed in the previous section. Thus, on the basis of the EPRI Test Program, it is concluded that the operability of the FBNP safety valves is confirmed.
As has been noted in the EPRI Test Program reports, valve flutter occurred during some tests with loop seals; however, subsequent valve performance was not affected substantially. Also, delayed lift (until the loop seal had cleared) and instances of valves opening and closing slightly outside the system specifications were observed on some tests. The impact of these test valve response characteristics on PBNP is discussed below.
Valve Flutter On certain of the EPRI tests, the valve stem did iio:
immediately open to its rated, full-lift position.
Rather, oscillation (flutter) occurred, particularly during clearance of the solid water in the loop seal. In more severe instances, the valve reclosed during these oscillations (chatter). In other tests, this flutter (or chatter) occurred during the closing cycle.
The flutter and chatter was confined to long inlet configurations. It was most pronounced on tests in I which solid water conditions occurred. With the exception of the loop seal water clearance, these solid water conditions are not applicable to PBNP. j Note also that chatter only occurred on three of the tests listed in Table 3-2. These were all low ]
i backpressure tests - PBNP has high backpressure.
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Impell Report No. 09-0870-0014 Revision 1 3.0. POINT BEACH NUCLEAR ~ PLANT EVALUATION
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The phenomenon is apparently related to upstream water hammer. Upon initial valve lift, a pressure wave (pressure drop).is propagated upstream.through the long inlet piping. This wave is reflected at the pressurizer interface and travels back to the valve.
The resultant pressure drop at the valve.may cause the valve to momentarily start closing until the prsscure rebuilds. Under certain resonance conditions between the valve stem and this pressure wave, flutter may occur.
1 The tendency of flutter to occur will thus be ,
dependent on the valve characteristics, the fluid i conditions, and the length of the inlet piping - that is, the distance to the pressurizer.
In interpreting the significance of the EPRI Test Program results with respect to flutter and chatter, it is assumed that the valve characteristics of the PBNP-specific 4K26 valves and the 6H6 and 3K6 valves are essentially the same.
For the test on the 3K6 valve listed in Table 3-2 as directly applicable to PBNP, the valve response was stable. For the three corresponding tests on the 6M6 valve,'some flutter occurred in each case. .This was confined to the period during loop seal clearance.
As noted in the previous section however, the extent and volume of the loop seal and the length of the inlet piping at PBNP is considerably less than for either the 6M6 or 3K6 tests. Based on this , it is concluded that any flutter that may occur at PBNP will be of lesser extent and severity than that obrerved in these EPRI tests. This is because less lotp seal water must be discharged and subsequent water hammer will be of higher frequency and thus unlikely to resonate with the valve.
Furthermore, should flutter (or even chatter) occur, it will not affect the operability of the valves.
They will continue to open to relieve pressure and to close as-required. Damage to the valve seat may i l
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occur, but any post-activation leakage (the maximum observed in the tests listed in Table 3-1 was 1.5 gpm) will not be significant.
Delaved lift ~
In most tests on the loop seal configuration, the valve opened partially to discharge the loop seal water before opening fully on the subsequerit saturated steam. The time between initial opening and full lift was of the order of one second or less.
Similar behavior would be expected of the PBNP valves. However, given the lesser volume of the PBNP loop seals (370 cubic inches versus 470 and 1760 cubic inches for the 3K6 and 6M6, respectively), the time between initial opening and full lift will be correspondingly less.
This slight delay in full lift is not considered to be of consequence in evaluating the pressure relief system's ability to relieve the postulated operating transients for PBNP.
Valve Ooenino Pressure Valve specifications require that the safety valves open with 3 percent of their set pressure. The opening pressures of the 3K6 and 6H6 safety valves in the tests relevant to PBNP (Table 3.1) ranged to
+8.9% of the set pressure.
Given the conservatism of the limiting, locked rotor transient definition for PBNP - which, among other conservatism, does not consider the relieving capacity of the PORV's - the variation is not considered significant.
Valve lift and Flow The 3K6 generally achieved rated lift and at least 90 percent of the required rated flow. The 6H6 generally exceeded the required rated flow. Based on i
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Impell Report No. 09-0870-0014 Revision 1 3.0 POINT BEACH NUCLEAR PLANT EVALUATION Mn 44, a A s w. e.:. d this, it is deduced that the PBNP 4K26 valves will provide sufficient relieving capacity.
1 Valve Blowdown The blowdown specified for the PBNP valves is 5 percent. The actual blowdown in the EPRI tests ranged from 5 to 10 percent for the 6M6 valve, and from 17 to 20 percent for the 3K6.
Based on this, the blowdown for the PBNP safety valves may exceed 5 percent. This is not considered significant. Since PBNP is designed to accommodate losses of reactor coolant resulting from postulated openings in the reactor coolant system, it is clear that increased blowdown is not a safety concern.
Rino Settinas For the Crosby 6M6 tests listed in Table 3-1, a limited variation of the ring settings was carried out. The effect of these ring setting variations on Crosby valve performance during the tests was not significant.
3.3 Ihm.rmal-Hydraulic This section describes the development of force time
&nalysis histories induced on the piping system by safety valve actuation. This includes:
1 Development of a thermal-hydraulic computer j model of the system I Performance of analysis to determine transient state histories at discrete locations Integration of these transient state histories to develop force time histories on the piping 3.3.1 Thermal-H.ydraulic The thermal-hydraulic analysis was performed Models using the computer program REALPS/ MODI, which is described in Appendix A.
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I RELAP5/H001 thermal-hydraulic models were developed 1 for each unit. Each consists of a number of fluid j control volumes connected by flow paths or j junctions. These volumes extend through the piping system from the pressurizer to the PRT, and through the rupture disc to the containment.
The Unit 1 model contains 196 control volumes and 196 interconnecting junctions. The, Unit 2 model used 210 control volumes and 210 interconnecting junctions.
The' size of the volumes used was decreased in regions where the hydrodynamic behavior is expected to change more rapidly. In particular, the control volume size in the areas where loads could be underestimated due f to numerical smearing was maintained less than or !
equal to the loop seal water volume. Since the i probability of numerical smearing decreases as the ]
water slug travels downstream,-the control volume size was increased gradually toward the PRT.
Control junctions were included at all changes in .
flow area - such as the safety valves, reducers, i tees, and the pressurizer and PRT nozzles. Other junction locations were chosen to maintain dynamic stability and to provide sufficient force detail.
Individual control volumes and junctions were defined in terms of fluid state and phase parameters, geometry, and flow characteristics. The boundaries were placed to ensure adequate representation of the fluid transient.
To model each valve, a valve area, opening time, and loss coefficient were input. The critical flow correlations built into the code determine the valve flow rate based on these input parameters and the
- inlet pressure.
The alternate choking model in RELAP5/H001 was not i used in the discharge piping for the transient '
calculations. It was, however, used upstream of and at the valves since choked flow would occur in these areas.
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____________________.__J
Impell Report No. 09-0870-0014 Revision 1 3.0. POINT BEACH NUCLEAR PLANT EVALUATION wmewsua61 ku:a w nnm-mny&autumwauauM sax;&t.uM me:w mmmwn,m iuriswn~y4 3.3.2 Parameters and Assu- This section defines (and describes the basis for) motions for Thermal- key assumptions and parameters for the thermal-Hydraulic Analysis hydraulic analysis.
Safety Valve Parameters Each unit has two safety valve mounted on the pressurizer. Pertinent safety valve data is given in Table 2-1.
Flow Rate - The rated capacity of each Point Beach safety valve is 288,000 lbs/hr of saturated steam.
This value was used in the analyses. In the EPRI tests, the larger 6M6 valve achieved flows in excess of its rated capacity, while the smaller 3K6 achieved flows slightly lower than its rated capacity. Thus, the use of rated capacity is considered appropriate.
Set Pressure - Both safety valves were conservatively A assumed to open simultaneously with the PORV's at the fi\ ,
peak pressurizer pressure of 2763 psig. Actual ~
safety valve setpoint is 2485 psig.
Valve Ooenina Time - Valve opening (pop) time was based on the valve achieving full lift in 20 milliseconds. This is conservative. The poptimes recorded for the 6M6 and 3K6 valves in the EPRI test ranged from 20 to 80 milliseconds.
PORV Parameters Each unit has two PORV's. These are attached to the pressurizer through a common nozzle. Pertinent PORV-data is given in Table 2-2.
Flow Rate - The maximum rated capacity of each PORV is 210,000 lb/hr. This value was used in the analyses.
l l
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Set Pressure - The set pressure for the PORV's is 2334 psig; however, the analysis conservatively assumes that both valves open at 2763 psig. The
, degree of conservatism introduced by this assumption l has been investigated using a sequential valve i actuation model - the effects were'found to be j negligible. 1 Valve Ooenina Time - The thermal-hydraulic analysis assumes a 0.80 second time for the PORV's. j l- Initial Conditions li The initial conditions for components upstream of the safety and power-operated relief valves were assumed I
to be those of the pressurizer. The pressurizer was assumed to contain saturated steam. The initial conditions for downstream components were assumed to be those of the PRT, the normal operating pressure of which is less than 5 psig. Initially, it was assumed to contain water and nitgrogen at 14.7 psia.
Looo Seal Temperature Profile - The safety valve loop I seals were assumed to be heated by means of enclosed insulating boxes as shown in Figure 3-1 (see discussion in Section 1.3). The temperature profile l used in the RELAP5/H001 analysis is shown on Figure l 3-2.
1 Heat Structure Model Condensation effects were conservatively ignored because considerable uncertainties are involved in the definition of a heat transfer coefficient, and a leaky valve would cause high pipe wall temperatures, thereby reducing the beneficial effect of wall heat transfer.
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Impell Report No. 09-0870-0014 Revision 1 3.0 POINT BEACH NUCLEAR PLANT EVALUATION EESMEU22E3ElIEdId2522EI25fEYEEEEfEA PRT Level Under normal operating conditions, the PRT is 72 ;
percent full. This level was used in the analysis.
As the level only determines quench capacity, the
.short-duration transient considered in this analysis ;
would not be sensitive to differences in the tank level.
3.3.3 Develocatant of Force After the transient state histories were determined Time Histories using RELAP5/HODI, force time histories on the piping system at changes in flow direction and flow area were generated using REFORC. REFORC is described in Appendix A.
These forcing functions include wave forces (control volume forces) and blowdown forces (control surface forces). Gravity forces were determined separately within the piping analyses.
3.3.4 Thermal-Hydraulic From the RELAP5/M001 analyses, transient pressures in i Results and Discussion the piping system, backpressure on the safety i valves, and steady-state temperature profiles were {
determined. From the REFORC analyses, forces on the piping system were calculated.
The analyses were reviewed for reasonableness. In particular, the adequacy of the simulation of the valves was verified.
Key results for both units are summarized below. As expected, results for the two units are very similar
- only in the region of the PRT, where the geom tries differ significantly, do the results show marked differences.
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r-Impell Report No. 09-0870-0014 Revision 1 3.0 POINT BEACH NUCLEAR PLANT EVALUATION m, vn,, . v , ,- m -, ~g,ge n, g. .n e n ,-m u.5 % Ys&L.d.J (JO bd. ba}bNkSSie k > LS . N i J as5.knA Aun,a k$
Backoressures - A typical plot of safety valve backpressure is shown on Figure 3-3. Maximum backpressure for each valve are summarized on Table 3-3.
Temperature Profiles - Similarly, maximum discharge line temperatures were those in the steady-state phase at the end of the transient. Maximum temperatures at each valve and at the PRT nozzle are given on Table 3-4.
Discharae Pioina Forces - The force time histories on the elbow immediately downstream of safety valve PCV-435, on the second elbow downstream of PORV PCB-431, and on the fourth elbow from the PRT nozzle are shown in Figures 3-4 through 3-7.
3.4 Pioina Evaluation 3.4.1 Jurisdictional Limits The piping evaluated includes the upstream piping from the pressurizer outlet nozzles to the safety and ,
power-operated relief valves, and the downstream (or l discharge) piping from each of these valves to the j PRT nozzle. 1 The 4-inch branch lines from relief valves 1-314 and 2-314 (Unit I and 2, respectively), which join the 8-inch discharge header in the region of the PRT, ;
were modeled beyond these valves so as to account correctly for their influence on the discharge piping response.
l Loads on valves, nozzles, and flanges were determined, but no evaluation of the adequacy of these components was performed.
3.4.2 Mathematical Models The piping system for each unit was idealized as SUPERPIPE mathematical models. These consist of concentrated masses connected by massless elastic members. The concentrated masses were located so as to adequately represent the dynamic properties of the system.
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.Impe11 Report No. 09-0870-0014 Revision 1 3.0 POINT BEACH NUCLEAR PLANT EVALUATION pmec ne:, egg,eww mmmyp&&Adda rag yw s.awusn c.K:Mr &:sA1wa n.:,d w u +.h u m d ada,N.6,;m & k Impell's computer program SUPERPIPE was used for all analyses. SUPERPIPE performs static, dynamic response spectra, and transient dynamic analyses. It also performs the required load combinations, code verification, and support load summaries. A description of SUPERPIPE is included in Appendix A.
l The piping system supports were modeled by specifying I the support type and applicable support direction. 1 Actual support stiffnesses were calculated and included where appropriate. ;
1 3.4.3 Description of Analyses Deadweiaht Analysis The weight of the piping, components, and contained water (as appropriate) was applied. The design I preloads of the spring hangers were modeled as vertical forces on the pipe. Snubber supports were assumed inactive for this analysis.
Thermal Exoansion Analysis I l
l For the calculation of secondary stresses due to J thermal expansion, the following design temperatures were used:
Piping upstream of safety valves and PORV's - I 680*F l i
Balance of piping - 477'F The pressurizer was also assumed to be at 680*F. The )
stress-free temperature for the analysis was taken as 70*F. Neither spring hangers nor snubber supports were included in the thermal anaiysis.
Seismic Analysis The seismic analysis input was based on Reference 11.
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Impell Report No. 09-0870-0014 {
Revision 1 '
I 3.0 POINT BEACH NUCLEAR PLANT EVALUATION IE33315E17752sETdEC33EEEMEiklEI3CIM Two co-directional earthquakes were modeled (X + Y, and Z + Y). Each was treated as an independent event, and the envelope of the resulting response was used. Consistent with the design basis for PBNP, Differential seismic building movements were not considered.
The pressurizer was included in the mathematical model for the seismic analysis in order to accurately represent its seismic input to the piping system.
The spectral curve used in the analysis is shown on Figure 3-8. The spectral accelerations given by this curve are conservative in that they envelope the spectral curves for the different support levels in the piping system. The corresponding accelerations for the Maximum Potential Earthquake were obtained by doubling these Design Basis Earthquake values. ,
l Damping was taken to be one-half of one percent of I critical for all seismic analyses.
Valve Discharae Time History Analysis Thermal-hydraulic force time histories at changes in flow direction and flow area, calculated for each unit by REFORC, were applied.
The direct integration solution method was used.
SUPERPIPE allows the system dynamic characteristics to be written as a set of differential equations of the form:
Hu + Cu + Ku - P where H, C, and K represent the mass, damping, and l stiffness of the system, u is the time-dependent displacement, and P is the applied load.
This set of equations is solved in coupled format by generating the response of the system as a function of the response at the previous time step. By assuming that the damping matrix is a linear combination of the mass and stiffness matrices, two unique frequency damping ratio pairs can be selected.
21
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Impell Report No. 09-0870-0014 Revision 1 3.0 POINT BEACH NUCLEAR PLANT EVALUATION Dii2ESTi1M23EllE2?$MZEE2?$2[$3ELGE These values were taken as two percent of critical damping at both the fundamental structural frequency and at the highest significant mode considered in the analysis, 125 cycles /sec. The frequencies of interest - that is, those between these limits - are conservatively underdamped.
The integration time step for the time history analysis was selected to provide accurate response in the higher frequencies of the system. A value of one millisecond was'used.
The event durations were taken as 0.70 and 0.75 seconds for Units 1 knd 2, respectively. Stresses were determined using the maximum of each moment component.
During certain of the EPRI tests, very high frequency pressure spikes were recorded in the upstream, loop seal piping. These water hammer stresses occurred principally during valve opening and were assocated with valve flutter. For a discussion of the phenomenon, see Section 3.2.2.
These high frequency pressure spikes were not included in the time history analysis. The B31.1 code allowable stresses are based on quasi-statically applied pressure throughout the pipe, not on localized pulses. Furthermore, should these pressure i spikes actually occur at PBNP, they would be of even higher frequency than those observed in the EPRI tests (due to the small volume of the loop seal and the short length of the inlet piping - see Section 3.2.2). It is not considered feasible that any significant permanent strain would occur in the PBNP piping. Thus, the potential for these pressure spikes is not of significance with regard to the piping integri ty. I 22 p,~L:,w$p- v , e m- -nv x,y, -
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Impell Report No. 09-0870-0014 Revision 1 3.0 POINT BEACH NUCLEAR PLANT EVALUATION r,, , m . m. .. ym m m ,- , m. _, . .m o t .-a , s.
- ..> . -.za.ma xa
...,y..,3.- .
a..awuw,s 3.4.4 Lead Combinations Load combinations are given in Tables 3-5 and 3-6.
The grouping method (modes with frequencies within 10 percent being regarded as closely-spaced) was used to combine modal components in the seismic analyses.
Seismic responses from multi-directional input were '
combined using the SRSS method. Seismic and valve l actuation responses were also combined by the SRSS I method.
For all pipe support combinations, the loads were maximized (maximum positive and maximum negative) by l considering the line both hot (thermal loads included) and cold (thermal loads not included).
3.4.5 Code Evaluation The Codp of Record for PBNP is USAS B31.1 (1967).'2 As part of the piping evaluation, the pipe stresses resulting from the above load
, combinations were compared to the appropriate I allowables.
1 In this evaluation, the (more conservative) stress intensification factors (SIF's) defined in ANSI B31.1 (1973)13 were used for piping design with the following exception: For butt-welded reducers, the Code of Record SIF was ad:pted.I4 ,
{
Allowable Stresses The piping between the pressurizer outlet nozzles and the safety and power-operated relief valves is seismic class piping. The discharge piping between these valves and the PRT is non-seismic class piping, ,
The allowable stresses for the seismic and non-seismic class piping are given on Tables 3-7 and 3-8, respectively.
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i Impe11 Report No. 09-0870-0014 Revision 1 3.0 POINT BEACH NUCLEAR PLANT EVALUATION wym .,e y., .
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lu -ix., w a-s c w , - - a a > .awa For the seismic class, pressure-retaining piping, the allowable stresses for load combinations 1, 2, 4, 6, and 6a Record.gre as per the FFDSARD and the Code of 82 The allowables for the valve actuation load combinations (numbers 3 and 5), which are not addressed in the FFDSAR, are consistent with the 1980 ASME Boiler and Pressure Vessel Code.lb In particular, the stress criterion for load combination 3 is the Level C (Emergency) service limit and that for load combination 5 is the Level D (Faulted) service limit.
The non-seismic class, non-pressure retaining discharge piping's function for the dynamic load cases (seismic and valve actuation) is to ' support' the valves and seismic class piping. -ihus, less restrictive allowables are appropriate.
To ensure that discharge piping integrity is t
maintained allowable perunder these tae 1980 ASME conditiong$
Code' (he 2.4 S )faulted was h
used.
Note that higher stress or nonlinear strain allowables may be appropriate for this non-seismic piping, with justification being provided that integrity is maintained such that the response of the seismic class piping and valves is not affected adversely.
24
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Impeil Roport No. 09-0870-0014 Revision 1,):a-
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4.1 Introduction Information obttined during. thehiping evaluation i describsd in Section 3.4 ns useA as a basis for detonpining possible concaptual inpport schemes.
Various support schemes were analyzed to determine tt e cot 1md configuration. The optimal configuration minimized the amount c(modification required to existinj inpports while k,eeping the number of new supports to a minimumi ,
- [, e t 'i,>
The[pipingevaluation'describedinSection3.4 1
4.2 Methodoloav -
in6 cated that t m4 cr,ty of the high pipe 1 stresses
'were due to the' valve discharge. time history , i
', analysis . Therefore, the valve discharge time i g histerh analysis was performed on various slipport y-' '
contifw.ations until the calcuiated pip'e stresses
, ,were %ithin allowable limits. At the same time, consideration was given to the to:ation of new supports using available field data' and the estimated
' thpacity of existing suppo,rts. ; .
Upon final!zation of the conceptual configuration, a full piping analysis waJ performed as described in Section 3.4 3. ,
t
/.
/ Load combination and code evaluation was performod as was done fo- the piping evaluation described in
,/ Sections 3.'4.4 and 3.4.5, respectively.
/ I! .
Load combiitations; are givenlin Tables 3-E and 3-6. t
,' All w able str' esses for the seismic and non-seismic vJtons n e'given in tam es 3-7 and 3-3 respectively.
i ,
4.3 Pioina Modifiu tioni Unit 1 - Gtfgitally,,tne Unit 1 piping system
,~
. contained 17 stiporn. For the final support
' configuration, thite of: these supports were deleted, th supports were modified and eight supports were qualified to the new loading conditions. Supports y' requiring modification were assigned an Impell mark f number and the appropriate calculations and drawings
, were volerated by Impell to perform the e modif cations. In additoa Ic the qualified and
- modified suports, eight supports were added to the Unit'l piplag system. Impell assigned mark numbers
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Impell Report No. 09-0870-0014 Revision 1
'4.0 POINT BEACH NUCLEAR PLANT MODIFICATIONS 12EE$GTSE!35E2 CIA 725E32525 Eld!2IC2EilIIIC to.the new supports and generated the necessary calculations and drawings to construct these additional supports. ,
A summary of the Unit 1 support configuration is given in Table 4-1.
Unit 2 - The Unit 2 piping system originally contained 16 supports. For the final support configuration, three of these supports were deleted, four supports were modified and nine supports were qualified to the new loading conditions. Supports requiring modification were assigned an Impell mark number and appropriate calculations and drawings were generated by Impell to perform the modifications. In addition to the qualified and modified supports, nine new supports were added to the Unit 2 piping system.
Impell assigned mark nurtbers to the new supports and generated the necessary calculations and drawings to construct these additional supports.
A summary of the Unit 2 support configuration is given in Table 4-2.
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Impel 1 Report No-.'_09-0870-0014 Revision 1 5.0 RESULTS-W EE EE M I C GI21TE E M & M E K EEE2 52ET E5EE
-5.1 Pioina Stresiel For full computer summaries, see Appendix C.
- Maximum stresses for each load combination are given in Tables 5-1 and 5-2.
The stresses are given for two regions of piping -
the seismic section and the non-seismic section.
The seismic section includes the inlet piping to the PORV's and their discharge piping to the downstream flanges located before the tee-junction. The seismic section also includes the inlet piping to the safety valves.
The non-seismic piping consists of the remaining piping.
5.2 Nozzle and Flance Detailed nozzle and flange stresses are included l
LOAdi in the computer summaries (see Appendix C). Table 5-3 gives the maximum components of load on each.
l l
5.5 Valve Accelerations Maximum safety valve accelerations in the horizontal and vertical directions for the modified piping systems are given on Table 5-4. These accelerations are due to valve actuation only.
5.4 Sucoort Loads For detailed suppport load summaries, see Appendix C.
Maximum support. loads from the load combinations given in Table 3-6 are shown on Table 5-5 and 5-6.
5.5 Discussion of Pioina Stresses Eioina Results Unit _1 - Seismic Class Pit,ina - Pipe stresses for this section of piping are within code allowables for load combinations i through 6a. The maximum ratio of actual stress divided by allowable stress is .92 for this section of piping.
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Impell Report No. 09-0870-0014 Revision i s
5.0 RESULTS EEXI'ZIT.TI.IIERT 7EfL3fEr5EErfl25IEFZE73 A summary of maximum stresses _ for this section of piping is shown in Table 5-1.
For full computer stress summaries, refer to Appendix C.
Unit 1 - Non-Seismic Class Pioing - Pipe stresses for this section of the discharge piping system are within code allowables for load combinations 1, 5, 6 and 6a. The maximum ratio of actual stress divided by allowable stress is .96.
A summary of m6ximum stresses for this section of l piping is given in Table 5-1. l, For full computer stress summaries, refer to Appendix C. !
Unit 2 - Seitmic Class Pioina - Pipe stresses for l this section of piping are within code allowables for l load combinations 1 through 6a except for load !
combination 6. The code permits load combination 6 ]
allowables to be exceeded as long as load combination '
6a allowables are satisfied. This requirement is satisfied for this section of piping. Therefore, this section of piping is within code allowables.
The maximum ratio of actual stress divided by allowable stress for load combinations 1 through 5 and 6a is .96.
A summary of maximum stresses for this section of piping is given in Table 5-2.
For full computer stress summaries, refer to Appendix C.
l Unit 2 - Non-Seismic Cias3 Pioina - Pipe strestes for this section of the discharge piping system are within code allowables for load combination 1 and 5. .
The maximum ratio of actual stress divided by l allowable stress for load combination 1 and 5 is
.63. High thermal stresses occur at the bottom of the system near the pressure relief tank. This p m ; n ;am .>mlO ayyr m 6: g n .;:n%
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Impell Report No. 09-0870-0014 Revision 1
$$3$EIOfECEEE52ES2E2EN35UAE!E2$
5.0 RESULTS thermal stress is responsible for a calculated overstress of 1 percent for load combination 6a.
This calculated overstress is conservative because:
Design rather than operating temperatures were assumed.
The analysis assuroes no gap between active thermal supports and the piping - in fact, finite gaps exist and this will decrease the thermal stress.
i The region of overstresss is outside the pressurizer cubical and far downstream of the pressurizer relief valves. Therefore, using the same design temperature for this portion of the line as is used near the pressurizer is conservative.
Therefore, while the calculated ctresses for loading combination 6a result in a 1 percent overstress, the actual stresses will be much lower and well within code allowables.
A summary of maximum stresses for this section of the discharge piping system are given in Table 5-2.
For full computer stress summaries, refer to Appendix C.
SuoDort Loads In general, original supports active during dynamic loadings were subjected to loads in excess of their :
original design loads. Original supports were either qualified or modified to support the additional loads due to valve actuation. Existing supports determined unnecessary by the piping analysis were removed from the system to reduce maintenance and inspection requirements. New supports were added to the system as required by the piping analysis.
A summary of the Unit 1 and Unit 2 support configurations are shown in Tables 4-1 and 4-2 respectively.
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Impell Report No. 09-0870-0014
[EfEIG52EEEI2TEETMIIISEED213Gli"1139' Revision 1 REFERENCES ~
- 1. "EPRI PWR Safety and Relief Valve Test Program Safety and Relief Valve Test Report" (Interim Report), Electric Power Research Institute, dated April 1982.
- 2. RELAPS/ MODI l Code Manual, NUREG/CR-1826.
- 3. REFORC V.2A: A Computer Program for-Calculating Fluid Forces Based on RELAPS Results, User's Manual, Revision .1, dated June 1982.
- 4. SUPERPIPE Users Manual, EDS Nuclear Inc, Version 15C, dated June 25, 1982.
- 5. Final Facility Description and Safety Analysis Report - Point Beach Nuclear Plant Unit No. I and 2," Chapter 14, Histcnsin Electric Power Company and Hisconsin Michigan Power Company.
- 6. " Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valves in Westinghouse-Designed Plants" (Interim Report), Westinghouse Electric Corporation, dated February 1982.
- 7. "EPRI/CE Safety Valve Test Data for the Crosby 3K6 Safety Valve (Long Inlet Pipe Configuration)," Electric Power Research Institute, November 10, 1981.
- 8. "EPRI/CE Safety Valve Test Data for the Crosby 6M6 Safety Valve (Long Inlet Pipe Configuration)," Electric Power Research Institute, January 12, 1982.
- 9. EPRI/CE Safety Valve Test Data for the Crosby 6H6 Safety Valve (Long Inlet Pipe Configuration)," Electric Power Research Institute, February 18, 1982.
I
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T
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L REFERENCES l
- 10. "EPRI PHR Safety and Relief Valve Test Program Valve Selection / Justification Report" (Interim l
Report), Electric Power Research Institute, 4 dated December 1981. 1 t
- 11. " Seismic Analysis - Point Beach Nuclear Plant Units One and Two Reactor Building Job No.
6118", Bechtel Corporation, dated March 1970. ,
i
- 12. USA Standard B31.1.0-1974, " Power Piping," )
American Society of Mechanical Engineers. ;
i
- 13. American National Standard ANSI B31.1-1973,
" Power Piping," American Society of Mechanical Engineers,
- 14. Letter from Hisconsin Electric Power Company to EDS Nuclear., " Point Beach Nuclear Plant Stress Intensification Factors," dated November 22, 1982.
- 15. "1980 ASME Boiler and Pressure Vessel Code." an American National Standard, American Society of Mechanical Engineers.
- 16. " Evaluation of Pressurizer Safety and Relief Valve System for Point Beach Nuclear Plant" EDS Report No. 09-0870-005-001, Revision 0.
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Impe11 Report No. 09-0870-0014:
Revision 1
' Tables 2'-1: Safetv Valve Parameters Number of Valves (per unit) 2.
Manufacturer Crosby Valve and Gage Co.
Type Spring-loaded nozzle type
- Designation: Size 4 K2 6 Style HB-BP-86 Type- E Height 520 lbs.
Steam Flow Capacity 288.000 lbs/hr (sat. steam)
(rated and maximum)-
Inlet Outlet ~
Design Pressure and 2485 psig. 500 psig Temperature 650*F 470*F Set Pressure- 2485 psig Table 2-2': Power-Ocerated Relief and Block Valve Parameters Number of Valves (per unit) PORV's: 2 Block: 2 Manufacturer PORV's: Copes-Vulcan Inc.
Block: Ve.an Type PORV's: Globe Valves Block: Gate Valves Steam Flow (PORV's) 210,000 lbs/hr (max) 179,000 lbs/hr (normal)
Design Pressure and 2485 psig/650*F Temperature (PORV's and Block Valves)
Set Pressure (PORV's) 2335 psig
p; .
Impell Report No. 09-0870-0014'
, Revision 1
. Table 3-1: . Applicable EPRI Tests.
for PBNP Safety Valves' 3K6 Valve 6M6 Valve ,
Tests Tests 525 906 526 908 529 910 536 913 917' 920*
923 929 1406 1415*
1419*
- heated loop seal Notes:
- 1. Above tests are for a filled loop seal
- 2. . Test fluid is saturated steam
Impell Report No. 09-0870-0014 Revision 1 Tables 3-2: Comparison of Results for Aeolicable EPRI Tests i
Directlyl EPRI Test Pressure Ramp Steady State Applicable Number Rate (osi/sec) Backpressure (Dsia) to PBNP 1
(3K6 Valve) !
525 3 445 526 200 520 529 18 385 536 8 432 <
(6M6 Valve) 906 3 253 908 297 613 910 375 227 l 1
913 375 233 917 291 238 920 297 240 923 283 650 929 319 700
- 1406 325 245 1415 360 245 1419 360 240 Nats:
- 1. That is, both high backpressure and high pressure ramp rate.
_ _ _ _ _ _ _ - - - _ - - - - - - - - - - 1
l 1
-Impell Report No.109-0870-0014 Revision 1 Table 3-3: Maximum Calculated Backeressures
' Backpressure (osial Safety Valve PCV-434 635 Safety Valve PCV-435 546
-Unit 2 Safety Valve PCV-434 586 Safety Valve PCV-435 555 Table 3-4: tiaximum Calculated Temperatures Unit 1 Temperature (*F)
Safety Valve PCV-434 682 Safety Valve PCV-435 682 PRT Nozzle 364 Unit 2 Safety Valve PCV-434 682 Safety Valve PCV-435 682 l PRT Nozzle 364 4
1
Impell Report.No. 09-0870-0014-R_ev.ision 1 Table 3-5: . Load Combinations for Pioina Analysis-Load Combination Number- Load Combination 1 (Sustained) Pr + Gr 2 (Occasional) Pr + Gr + OBE 3 (Occasional) Pr + Gr + SOT 4 (Faulted) Pr + Gr + SSE 5 (Faulted) Pr + Gr + SSE + SOT 6 (Thermal Expansion) Th-6a (Thermal Expansion Th + Pr + Gr and Sustained)
Table 3-6: Pipe Support Load Combinations Combination Loading Support Design Number Condition Load 1 . Sustained Gr + Th 2 Occasional Gr + Th + OBE 3 Occasional Gr + Th + OBE + SOT 4 Faulted Gr + Th + SSE 5 Faulted Gr + Th + SSE + SOT wJiara Pr - Pressure Gr - Gravity OBE - Design Basis Earthquake L '
SOT - System Operating Transient (Valve Actuation)
Th - Thermal SSE - Safe Shutdown Earthquake
_ -b
Impell' Report No.- 09-0870-0014
-Revision 1
)
... Table 3-7: Allowable Stresses for. 1 g , Seismic Class Picina Load Combination Allowable- I Number Streess l' (Sustained) 1.0 Sh 2 -(Occasional) 1.2 Sh 3 (Occasional) 1.8 Sh 4 (Faulted). 1.8 Sh '
5 (Faulted) 2.4 Sh 6 (Thermal Expansion)- SA 6a (Thermal Expansion SA+Sh and Sustained)
Table 3-8: Allowable Stresses for Non-Seismic Class Pinina Load Combination Allowable Number Stress 1 (Sustained) 1.0 Sh ,
l 5 (Faulted) 2.4 Sh 6 (Thermal Expansion)- SA L 6a (Thermal Expansion SA+Sh and Sustained) ,
i where SA - (1.25 Sc + 0.25 Sh )
!!Q1e: .l
.The allowable stress for load )
combinations number 2, 3, and 4 i is 2.4 Sh - these cases are thus I enveloped by load combination 5. !
1
Impell Report No. 09-0870-0014 Revision 1 Table 4-1: Unit 1 Succort Configuration ORIGINAL SUPPORTS FINAL SUPPORT CONFIGURATION Support Data Support Required Support Data Support Mark No. Pt. No. Tvoe Action Mark No. Pt. No. Tvoe HS-14 15 Snubber Modified HS-2501R-15 15 Snubber RC-16 21 Spring Qualified RC-16 21 Spring RC-18 31 Spring Qualified RC-18 31 Spring S-247 37 Rigid Hodified RS-601R-37 37 Rigid l HS-14 43 Snubber Modi fied HS-2501R-43 43 Snubber RC-14 C88 Spring Quali fied RC-14 C8B Spring RC-15 50 Spring Quali fied RC-15 50 Spring RS-200 71 Rigid Deleted l HS-17 73 Snubber Modified HS-601R-73 73 Snubber RC-17 74 Spring Quali fied RC-17 74 Spring HS-18 80 Snubber Modified HS-601R-80 80 Snubber l RC-12 82 Spring Qualified RC-12 82 Spring HS-200 84 Snubber Deleted
- S-248 86 Rigid Modified RS-601R-86 86 Rigid i RC-13 91 Spring Deleted H-200 93 Rigid Quali fied H-200 93 Rigid RC-13A 95 Spring Qualified RC-13A 95 Spring Additional Supoorts Support Data Support Mark No. Pt. No. Tvoe - - -
HS-2501R-22A 22A Snubber HS-2501R-22A 22A Snubber HS-601R-37A 37A Snubber HS-601R-37A 37A Snubber HS-2501R-51 51 Snubber HS-2501R-51 51 Snubber RS-601R-85 85 Rigid RS-601R-85 85 Rigid RS-601R-89A 89A Rigid RS-601R-89A 89A Rigid HS-601R-90 90 Snubber HS-601R-90 90 Snubber RS-601R-92 92 Rigid RS-601R-92 92 Rigid HS-601R-92A 92A Snubber HS-601R-92A 92A Snubber Notes:
- Support HS-200 was deleted from the final support configuration but can remain on the system without any adverse effects on the piping analysis
~
Impell Report No. 09-0870-0014 Revision 1 Table 4-2: Unit 2 Succort Configuration ORIGINAL ~ SUPPORTS FINAL SUPPORT CONFIGURATION Support Data Support -Required Support Data Support Mark No. Pt. No. Tvoe Action Mark No. Pt. No. Tvoe HS-28 15 Snubber Modi fi ed - HS-2501R-15 15 Snubber 2RC-14 21 Spring Qualified 2RC-14 21 Spring 2S-265 37 Rigid Modified RS-601R-36 36 Rigid Modified HS-28 43 Snubber HS-250lR-43 43 Snubber 2RC-15 44 Spring Qualified 2RC-15 44 Spring 2RC-13 50 Spring Qualified 2RC-13 50 Spring 2S-266 58 Rigid Deleted H-201 C14A Rigid Deleted-HS-30 72 Snubber Qualified HS-30 72 Snubber 2RC-10 73 Spring Qualified 2RC-10 73 Spring HS-29 80 Snubber Qualified HS-29 80 Snubber 2RC-11 81 Spring Qualified 2RC-ll 81 Spring HS-1 82 Saubber Deleted * ,
2S-265 85 Rigid Modi fied RS-60lR-85 85 Rigid ,
2RC-12 92 Spring Qualified 2RC-12 92 Spring
'H-200 99 Rigid Qualified H-200 99 Rigid Additional Sucoorts Support Data Support Mark No. Pt. No. Tvoe - - -
HS-2501R-21A 21A Snubber HS-2501R-21A 21A Snubber HS-601R-37 37 Snubber HS-601R-37 37 Snubber HS-2501R-49 49 Snubber HS-2501R-49 49 Snubber RS-60lR-83 83 Rigid RS-60lR-83 83 Rigid RS-601R-89A 89B Rigid RS-601R-898 898 Rigid RS-601R-91 91 Rigid RS-601R-91 91 Rigid RS-60lR-93 93 Snubber RS-601R-93 93 Snubber RS-601R-94 94 Rigid RS-60lR-94 94 Rigid HS-601R-95B 95B Snubber HS-60lR-958 958 Snubber Notes:
- Support HS-1 was deleted from the final support configuration but can remain on the system without any adverse effects on the piping analysis.
I
- - ;, , 1 tx ,
Impell. Report No.-09-0870-0014
' Revision 1 Table 5-1: Unit 1 Maximum Ploe Stresses Load Maximum A11owable2 Combination Joint Stress, Stress,
. Number Name (osi) (osi)
. SEISMIC CLASS SECTION 1 9 11,080 16,000- l 2 10 15,472 19,200 3 73 23,689 28,800 4 10 20,653 28,800-5 73 35,651 38,400 6 . 236 24,194 27,438 6a 236 29,463' 43,438 NON-SEISHIC CLASS SECTION ,
1 117 5,313 14,642 5 181 23,017 35,141 6 117 25,891 27,098 6a 117 31,204 41,740 I
Notes:
- 1. For Joint Name/ Location, see Appendix 8.
- 2. Per Tables 3-7 and 3-8 .
1
Impell Report No. 09-0870-0014' Revision if
. Table 5-2: Unit 2 Maximum Pice Stresses Load Maximum- A11owable2 Combination Joint. Stress, Stress, Number Name (osi)- (osi)
SEISMIC' CLASS'SECTION' 1 9 11,502 16,000~
i 2 37 16,601 19,200
-3 '87 22,321 28,800 4 37 26,041 28,800 5 87' 36,928 -38,400 t
6 236 27,520 27,438 6a. 236 33,340 43,438 l
1 UpN-SEISHIC CLASS SECTION 1 223 7,535 14,642 5 92 22,244 35,141 6 184 38,904 27,098 6a 184 42,405 41,740 i
Hatn:
- 1. For Joint Name/ Location, see Appendix 8.
- 2. Per Tables 3-7 and 3-8
E ~Impell Report No.. 09-0870-0014:
Revision 1, Table 5-3: Nozzle /Flance Loads Axial Resultant Torsional Bendina Moment Nozzle / Load Shear Load. Homent
. Flange Load Case ilhil (1bs) (ft-lbs) Mv(ft-Ibs)_Hz(ft-lbs) l i dall_1 E98Y Gravity 87 81 160 132 368 MazIle Thermal 101 1320 1397 1904 707 at inter- Seismic OBE (X+Y) 107 S1 106 155 200 face of Seismic OBE (Z+Y) 226 158 149 415 534 press. Seismic SSE (X+Y) 214 162 212 310 400 nozzzle & Seismic SSE (Z+Y) 452 315- 298 830 1068 PORV SOT 1034 1313 1697 1307 1849
' inlet piping SE Pioina Gravity 388- 417 94 121 115 L Nozzle Thermal 3684 1225 2434 '3890 3543 at inter- Seismic OBE (X+Y) 81 87 48 39 67 l- face of Seismic OBE (Z+Y) 79 91 66 60 82 I
press. Seismic SSE (X+Y) 162 174 96 78 134' nozzzle & Seismic SSE (Z+Y) 158 183 132 120 164 PCV-434 SOT 1644 1638 1741 1754 2574 inlet
-piping SV- Gravity 408 431 19 21 189 Pioina Thermal 443 802 168 638 874 Ngzzlf Seismic OBE (X+Y) 105 81 45 34 169 at inter- Seismic OBE (Z+Y) 100 105 35 30 222 face of Seismic SSE (X+Y) 210 163 90 68 338 press. Seismic SSE (Z+Y) 200 210 70 60 444 nozzle & SOT 2000 1759 1231 1028 2449 PCV-435 inlet piping ERI Gravity 360 163 237 184 27 at Thermal 1366 588 1800 2192 5654 flange Seismic OBE (X+Y) 726 493 608 1058 309 Seismic OBE (Z+Y) 327 533 271 2494 253 Seismic SSE (X+Y) 1452 987 .216 2116 618 Scismic SSE (Z+Y) 654 1065 542 4988 510 SOT 24193 10727 11949 5297 11099
Impoll Report No. 09-0870-0014
,. Revision 1
[
Table 5-3: '(con't) 1 Axial Resultant Torsional 8endino Moment Nozzle / Load Shear Load Moment Flanae. (pad Case ilhil _(1bs) (ft-lbs) Mv(ft-Ibs) MZ(ft-lbsl Elance Gravity 8 343 138 30 143 down- Thermal 332 756 345 -319 97
~ stream Seismic OBE (X+Y) 79 107 92 29 132 of PORV Seismic OBE (Z+Y) 224 148 '326 64 452 '!
1-PCV - Seismic SSE (X+Y) 158 213 184 58 264 431C Seismic SSE (Z+Y) 448 296 652 128 904 SOT 5618 2395 572 470 1449 fi1LMyle Gravity .
1 '130 61 25 152 I down- Thermal 378 339- 148 663 456 stream Seismic OBE (X+Y) 100 76 69 27 65 of PORV Seismic OBE (Z+Y) 96 150 164 89 130 1-PCV-430 Seismic SSE (X+Y) 200 152 138 54 130 Seismic SSE (Z+Y) 192 300 328 178 260
, SOT 6006 979 571 748 623 ECV-435 Gravity 346 17 20 199 549 at valve Thermal 205 893 397 1014 1809 inlet Seismic OBE (X+Y) 93 40 67 57 47 Seismic 08E (Z+Y) 97 51 88 59 46 Seismic SSE (X+Y) 186 81 134 114 94 Seismic SSE (Z+Y) 194 102 176 118 92 SOT 2800 2991 1787 1215 1662 ECy:435 Gravity 1 261 187 36 333 at valve Thermal 605 688 1511 .121 2431 outlet Seismic OBE (X+Y) 117 78 98 27 55 Seismic OBE (Z+Y) 81 91 103 26 70 Seismic SSE (X+Y) 234 156 196 54 110 Seismic SSE (Z+Y) 162 182 206 52 140 '
SOT 18648 2651 1870 1859 1868 PCV-434 Gravity 316 54 14 235 432 )
at valve Thermal 3356 1952 1003 3112 202 inlet Seismic OBE (X+Y) 85 33 16 37 56 Seismic OBE (Z+Y) 83 43 19 33 52 ,
Seismic SSE (X+Y) 170 67 32 74 112 ,
I Seismic SSE (Z+Y) 166 87 38 66 104 SOT 2469 3171 1511 1499 2475 ECY-414 Gravity 30 296 248 24 117 at valve Thermal 631 3831 4805 2543 1532 !
outlet Selsmic OBE (X+Y) 84 94 56 74 50 Seismic OBE (Z+Y) 45 103 61 80 48 Seismic SSE (X+Y) 168 187 112 148 100 Seismic SSE (Z+Y) 90 207 122 160 96 S0T 17426 4949 3252 3138 3179
Impell Report No. 09-0870-0014 Retfision 1 Table 5-3: (con't)
Axial Resultant Torsional Bendina Moment Nozzle / Load Shear Load Moment -
Flanae Lead Case (1bs) (lbs) (ft-1bs) Mv(ft-lbs) Mz(ft-1bs) !
flangt Gravity 115 304 262 284 257 on branch Thermal 27 58 70 91 143 i line up- Seismic OBE (X+Y) 425 785 2835 2169 2087 !
steam Seismic OBE (Z+Y) 675. 486 1456 959 2103 1 of PRT Seismic SSE (X+Y) 850 1571 5670 4338 4174 Seismic SSE (Z+Y) 1350 972 2912 1918 4206 SOT 6404 1218 464 1217 1400 Unit 2 fMLRY Gravity 222 235 145 197 406 3
!!przle Thermal 172 346 1071 560 349 at inter- Seismic OBE (X+Y) 91 269 214 361 140 "!
face of Seismic OBE (Z+Y) 76 399 229 -620 228 press. Seismic SSE (X+Y) 182 538 428 722 280 nozzzle & Seismic SSE (Z+Y) 152 798 458 1240 456
.PORV SOT 965 1131 1034 1945 1000 inlet piping SY Gravity 323 327 12 12 247 P_loina Thermal 199 1304 484 1013 782 HQZzig Seismic OBE (X+Y) 83 103 53 37 141 at inter- Seismic OBE (Z+Y) 95 121 40 34 170 face of Seismic SSE (X+Y) 166 206 106 74 282 press. Seismic SSE (Z+Y) 190 242 80 68 380 nozzle & S0T 2088 1619 1020 1324 3241 PCV-435 inlet piping SV Gravity 330 354 266 257 138 Pinina Thermal 3382 1264 2804 4185 2905 Enzzle Seismic OBE (X+Y) 67 81 54 54 63 at inter- Seismic OBE (Z+Y) 74 95 67 64 81 face of Seismic SSE (X+Y) 134 161 128 108 126 press. Seismic SSE (2+Y) 148 189 134 128 162 nozzle & SOT 1512 1471 1470 1375 1902 PCV-434 inlet piping i
Impe11 Report No. 09-0870-0014 Revision 1 Table 5-3: (con't)
Axial Resultant Torsional Bendino Moment Nozzle / Load Shear Load Moment Flange Load Case 11h11 (1bs) (ft-lbs) Hv(ft-lbs) Mz(ft-lbs)
ERI Gravity 793 402 202 200 738
'at Thermal 3460 85 1262 250 3152 flange Seismic OBE (X+Y) 31 158 337 487 53 Seismic OBE (Z+Y) 33 328 713 1032 95 Seismic SSE (X+Y) 62 316 674 974 106
. Seismic SSE (Z+Y) 66 655 1426 2064 109' SOT 18144 11736 8909 1485 10711 Flange- Gravity 11 394 69 41 73 down- Thermal 128 153 182 81 122 stream Seismic OBE (X+Y) 131 382 223 102 219 of PORV Seismic OBE (Z+Y) 203 .330 218 100 383 2-PCV- Seismic SSE (X+Y) 262 763 446 204 438 431C Seismic SSE (Z+Y) 406 659 436 200 766 SOT 5671 807 363 285 358 F1ance Gravity 1 31 75 10 518 down- . Thermal 100 160 153 322 201 stream Seismic OBE (X+Y) 132 194 269 281 433 of PORV Seismic OBE (Z+Y) 117 184 155 222 450 2-PCV- Seismic SSE (X+Y) 264 389 538 562 866 430 Seismic SSE (Z+Y) 234 368 310 444 900 SOT 6254 1295 462 408 633 PCV-435 Gravity 209 69 30 175 513 at valve Thermal 2 1319 387 910 1764 inlet Seismic OBE (X+Y) 97 45 57 52 37 Seismic OBE (Z+Y) 117 50 68 62 42 Seismic SSE (X+Y) 194 89 114 104 74 Seismic SSE (Z+Y) 234 100 136 124 84 l SOT 2838 1689 1447 1432 1469 l
PCV-435 Gravity 57 354 205 3 307 at valve Thermal 964 900 1595 363 2495 outlet Seismic OBE (X+Y) 95 76 64 34 85 Seismic OBE (Z+Y) 80 86 74 36 101 Seismic SSE (X+Y) 190 152 128 68 170 Seismic SSE (Z+Y) 160 172 148 72 202 S0T 15995 2684 2312 1637 1663 l 1
i l
Impe11 Report No. 09-0870-0014 ,
Revision 1 !
Table 5-3: (con't)
Axial Resultant Torsional Bendina Homent J Nozzle / Load Shear Load Homent l Flange Load Case 11bi). (1bs) (ft-lbs) Hv(ft-lbs) Hz(ft-lbs1 j PCV-434- Gravity 223 128 47 208 432 at valve Thermal 3017 1983 950 2717 91 inlet Seismic OBE (X+Y) 65 46 17 27 48 Seismic 08E (Z+Y) 79 56 20 30 54 ,
Seismic SSE (X+Y) 130 93 34 54 96 Seismic SSE (Z+Y) 158 112 40 60 108 SOT 2850 2923 1005 843 1994
)
ECV-435 Gravity 118 340 172 9 104 at valve Thermal 166 3607 4216 2604 2735 outlet Seismic OBE (X+Y) 80 75 39 50 49 Seismic 08E (Z+Y) 53 94 45 61 58 Seismic SSE (X+Y) 160 150 78 100 98 Seismic SSE (Z+Y) 106 188 90 122 116 S0T 12494 2556 1823 1617 4971 l Flange Gravity 52 271 49 355 704 on branch Thermal 109 451 153 390 2073 line up- Seismic OBE (X+Y) 57 137 298 536 227 steam Seismic 08E (Z+Y) 91 214 524 834 425 of PRT Seismic SSE (X+Y) 114 273 596 1072 454 Seismic SSE (Z+Y) 182 428 1048 1668 850 SOT 10081 3823 448 1781 8634
- 4 Impe11' Report No. 09-0870-0014 l Revision 1 l
' Table 5-4: Safety Valve Accelerations 1,2 Horizontal Vertical Safety Valve Acceleration ' Acceleration Unit 1 PCV-434 9.4 5.7 PCV-435 6.8 4.4 Unit 2 PCV-434 9.4 5.7 PCV-435 6.8 4.4 Notes:
~
- 1. The accelerations shown above are a result of valve actuation
- 2. All accelerations are in units of gravity (g's), and are given at the valve's center of gravity i
Impell Report No. 09-0870-0014
. Revision 1.
Table 5-5: Unit 1 Pioina Succort loads Support. Succort load (lbs)
. Mark No. Support Load Combination No.
(Data Point) Tvoe Direction 1 _2_ _1_ 4 _E_
HS-2501R-15 Snubber Fx -
213 2441 427 2654 (15)
RC-16 Spring Fy 750 750 750 750 750 (21)
'RC-18 Spring Fy 727 727 727 727 727 (31)
-RS-601R-37 Rigid Fx 1677 1803 3512 1929 3638.
(37) Fz 827 1311 2551 -1794 3035 HS-2501R-43 Snubber Fx. -
-105 1360 211 1464 (43)
RC-14 Spring Fy 325 325. 325 325 325
.(C8B)
RC-15 Spring Fy 501 501 501 501 501 (50)
HS-601R-73 Snubber Fx -
170 5969 340 6139
'(73)
RC-17 ~ Spring Fy 447 447 447 447 447 (74)
HS-601R-80 Snubber Fx -
97 4912 195 5009 (80).
RC-12 Spring Fy 1860 1860 1860 1860 1860 (82)
.," Impell Report No. 09-0870-0014-Revision 1 Table'5-5: Unit l' Pinina Suncort Loads l(cont'd)
L
. Support .
Suonort Load (1bs)
Mark No. . Support Load Combination No.
(Data Point)- Tvoe Direction 1 _2._ l_ 4 - 5_
RS-601R-86 Rigid Fx' 2603 2661 '4936 2718 4994 (86) Fz 3773. 4005 6018 4237 '6250 H-200 Rigid Fy 835 1426 3036 2017. 3627 (93) Fz .116 584 2815- 1052 3283 RC-13A . Spring Fy 386 386 386 386' 386 (95).
HS-2501R-22A Snubber Fy -
151 2178 302 2329 (22A)
.HS-601R-37A Snubber Fy -
54 1586 107 1640 (37A)
-HS-2501R-51 Snubber Fy -
228 1340 456 1568 (51)
RS-601R Rigid Fz 5633 5843 20583 6054 20794 (85)
RS-601R-89A Rigid Fx 1545 1589 5497 1634 5542 (89A) Fy 7926 8191 40576 8455 40841 '
Fz 522 622 1921 722 2021 HS-601R-90 Snubber Fx -
94 17180'- 187 17273 0 (90)
RS-601R-92 Rigid Fy 3722 3831 11713 3940 11823 l (92) I 1
HS-601R-92A Snubber Fx -
480 27232 960 27712 (92A)
Impe11 Report No. 09-0870-0014 Revision 1
~
Table 5-6: Unit 2 Pioino Sucoort Loads Support Suonort Load (1bs)
Mark No.- Support. Load Combination No.
.(Data Point) Tvoe Direction 1 _2_ _3_ 4 _5_
HS-250R Snubber Fx -
199 1031 399 1231 (15)'
2RC-14 Spring Fy 681 681 681 681 681 (21)
RS-601R-36 Rigid Fx 560 1114 2471 1668 3025 (36) Fz 398 1027 1741 1657 2370
-HS-2501R-43 Snubber Fx - 135 1205 271 1340 (43)-
2RC-15 Spring Fy 313 313 313 313 313 (44) 2RC-13 Spring Fy 684 684 '684 684 684 (50)
HS-S0 Snubber- Fx -
177 4274 353 4451 (72) 2RC-10 Spring Fy 261 261 261 261 261 (73)-
HS-29 Snubber Fx -
98 3689 197 3788 (80) 2RC-11 Spring Fy 3669 3669 3669 3669 3669 (81)
RS-601R-83 Rigid Fz 5500 5760 14451 6022 14713 (83)_
RS-601R-85 Rigid Fx 2056 2225 7242 2393 7411 (85) Fz 2930 3107 5242 3284 5420 2RC-12 Spring Fy 470 470 470 470 470 (92) '
i
e
[.
L L
Impe111 Report No.'09-0870-0014" Revision 1.
Table 5-6: (cont'd)-
Support Suonort Load '(1bs)
Mark No. Support Load Combination No.
(Data Poind Type Direction' i ._Z_ _3_ 4 _5._
H-200 Rigid Fx 7711 7856 '18997 8000 19141; (99) Fz 78 629 8651 -1179 9201 HS-2501R-21A Snubber Fy -
269 1068 537, 1337
'(21A)
HS-601R-37 Snubber Fy -
265 .2833 531 3098 (37)
HS-2501R-49 . Snubber- Fy ' -
213 1382 427 1595 (49)
RS-601R-89B Rigid Fx 715 755 6208 795 6248' (898) Fy 7653 7880 36425 8106 36651' RS-601R Rigid Fx 1408 1469 23291 1529 23352 (91)
RS-601R-93 -Snubber Fy -
60 9276~ 121 9336
.(93)
RS-601R-94 Rigid Fx 7458 7747 39718 '8036 40007 (94)
HS-601R-95B Snubber Fy -
120 19066 239 19185 (958)
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Impell Report No. 09-0870-0014 Revision 1 APPENDIX A: DESCRIPTION Qf_.10MPUTER PROGRAMS
)
SUPERPIPE l i
l SUPERPIPE is a comprehensive computer program developed by Impell for the ]
structural analysis and design checking of oiping systems. Analysis may be carried '
out in accordance with the requirements of any one of several standard piping codes.
SUPERPIPE executes in distinct phases; namely, specification of system geometry, .
static analysis, determination of dynaraic characteristics, response spectrum or l time history analysis, and design checking against code requirements. Appropriate '
combinations of these phases may be executed during any specific computer run.
SUPERPIPE can generate its own finite element mesh, lumped m sses being automatically positioned along the pipe.
Output from SUPERPIPE includes a detailed summary of stresses and displacements.
Results of analyses can be saved permanently on problem data files and recalled for use in subsequent computer runs. A code compliance summary based on any of several standard piping codes built into the program is output. Nozzle and penetration summaries are also available. SUPERPIPE features a number of post processors and plotting routines.
The SUPERPIPE program has been extensively benchmarked against several other piping analysis programs and has been found to be both accurate and cost-effective.
BELAPS/H001 RELAP5/H001 was originally developed to calculated PHER thermal-hydraulic loads induced by a loss-of-coolant accident. Recently, it has been benchmarked against the EPRI Safety and Relief Valve Test Program.
The basic parameters used in modeling the hydraulic network are control volumes and connecting junctions. RELAP5/M001 solves the conservation of momentum, energy, and mass equations for the resulting network of control volumes and junctions.
The program calculates thermal-hydraulic transient with a complete two-fluid, two-velocity, two-temperature description. A set of five equations (two mass, two momentum, one energy) describes the two fluids. The need for a second energy equation has been eliminated by assuming that the least-massive phase is at saturated conditions. Two-velocity phenomena such as entrainment and slip are A-1
l Impell Report No. 09-0870-0014 Revision 1 calculated by simultaneous solution of separate phasic mass and momentum equations. Interphase friction correlations are flow regime-dependent, and there is no reliance on direct-empirical correlations for slip velocity, flooding rate, or entrainment fraction.
Thermal nonequilibrium of either phase is accounted for in RELAP5/H001.
Calculations of evaporation / condensation determine the rate at which the two fluids reach equilibrium. One phase in each control volume is assumed to be at its saturated condition, thus, both subcooled water and superheated steam can be treated simultaneously in an overall model, but not within an individual control volume.
For liquid discharge, the critical flow rate is calculated in RELAP5/H001 by application of a modified Bernoulli equation between the upstream fluid volume and the choking plane. Nonequilibrium is accounted for by allowing the pressure at the choking plane to undershoot the local saturation pressure based on the Alamgir-Leinhard-Jones correlation. For two-phase discharges, the critical flow rate is calculated from a characteristic analysis of the conservation equations.
For vapor discharge, the critical flow rate is calculated based on the local fluid-sonic velocity.
REFORC REFORC was developed as part of the EPRI Safety / Relief Valve Test Program. It calculates the fluid forces acting on a piping network by application of Newton's Second Law of Motion.
The method of force-history generation is to develop the total transient force (Ft) in the axial direction at opposing components (such as bends or tees) according to the following equation:
Ft . Fw + Fcs Fw is the wave force due to the fluid acceleration and Fcs is the blowdown force due to the pressure and momentum at the control surface normal to the direction of F. Total transient forces are calculated in this fashion at variations in flow areas and/or changes in flow direction.
A-2
Impell Report No. 09-0870-0014 Revision 1 APPENDIX B: SUPERPIPE MODELS Models for Units 1 and 2 are held under separate cover l
I B-1 1
____.__________._____________.________J
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OVERSIZE DOCUMENT PAGE PULLED SEE APERTURE CARDS 1 NUMBER OFOVERSIZE PAGES FILMED ON APERTURE CARDS i
APERTURE CARD /HARD COPY AVAILABLE FROM RECORD SERVICES BRANCH,TIDC FTS 492 = 8989 O
Impell Report No. 09-0870-0014 Revision 1 APPENDIX C: DETAILED PIPE STRESS AND SUPPORT LOAD SUMMARIES Detailed summaries are held under separate cover.
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