ML20136F267
| ML20136F267 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 12/31/1985 |
| From: | Withers B PORTLAND GENERAL ELECTRIC CO. |
| To: | Varga S Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20136F274 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM TAC-44625, NUDOCS 8601070330 | |
| Download: ML20136F267 (19) | |
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Decemt;ar 31, 1985 But D Wens Vcc Pn;sdent Trojan Nuclear Plant Docket 50-344 License NPF-1 Director of Nuclear Reactor kegulation ATTN: Steven A. Varga Director, PWR-A Project Directorate No. 3 U.S. Nuclear Regulatory Commiision Washington DC 20555
Dear Mr. Varga:
Request for Additional Infotmation NUREG-0737 Iten II.D.1 By letter dated June 12, 1985, the NRC requested that Portland G69eral Electric Company provide additional information regarding peiformahep testing of relief and safety valves (NUREG-0737 Item II.O.1). Attached is Portland General Electric Company's response to this request.
An as-built evaluation of the modified Trojan Pressurizer Relief System has been pe'eformed, as first discussed in PGE-to-NRC letter dated July 12, 1983. The evaluation revealed n7 unexpected results, and the concitsions origina11.y presented in PGE-1039 have not (cianged.
If you have any questions, please advise.
Sincerely, r
Bart . Withertl Vice President Nuclear Attachment c: Regional Administrator, Region V U.S. Nuclear Regulatory Cottalssion Office of Executive Director for Operations gh
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1450 Maria Lane, Suite 210 I WaltYUt CrtOk CA 94596 Hr. Lynn Frank, Eirector State of Oregon SW.
Department of Energy u-2 nwrr (ne only.
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,' Trajcn Nuclo!r Picnt Mr. S. A. Varst Docket 50-344 Attachment License MPF-1 Decamber 31, 1985 Page 1 of 16 ADDITIONAL INFORMATIOW REGARDING NUREG-0737. ITEN TI.D.1
- . NBC Ouestion_i_ ,
The Trojan plaat specific submittal stated that the cold overprest re event was not representad by the Westinghoure 4-loop reference plant cold overpressure event. Provide additional detLil and discussion on how thL Trojait cold overpressur. inlet fluid conditions were determined.
PCs Response 1 TheTrejanNuclearPlantOverpressureMitigat$onSystem(OMS)makesuse of a design that e.tilizes the two precsurizer power-opereted relief
., valves with asiociatFd circuitry as a means of pressure relief during 4
cold overpressure events. The Westinghouse 4-loop reference plant also
- uses the two prer.surizer power-operated relief valves and associatbd '
circuitry. However, unlike Trojan, a variable relief pr~ essure is campt:tej by the circuitry based on Reactor Coolaut System (RCS) temperature. At Trojan, when the OMS is enabled, relief setpoints are sol at conctant values of 440 and 490 pai3 . For this tsason, the Westinghouse analyzed Plantm s pecific valve inlet fluid conditions depicted in Reference 1 resulting from cold overpressurization ovents are net applicable to Trojan. Operation of the OMS is contcolled by
, operating procedures. During ctartup from a solid Plant condition, a pressurizer bubble $s .4rown and then the OMS is disabled after RCS pressure is raised to 400 psig if all cold les temperatures cre above 290*F. During shutdoten, the OMS is onabled when RCS pressure is lowered to 375 psig or when any cold leg temperattre goes below 300*F. Then the pressurizer bubble is collapsed. Therefore, the maximum pressure and temperature the OMS normally sees is 400 psig and 300*F. Becacse of the i above operational conditiona, conditions of liquid, steam-liquid, and steam discharge may be seen at the julet to the pressurizer PORVs during i coJd ovorpressurization events.
, In the case of liquid discharge, References 2 and 3 transmitted to you
! the results of the analyses of the RCS rsspor se to possible pressure l transients that con occur during startup and shutdown. As part of those transmittals, copies of Feferances 4 and 5 were sent. The generic algorithms presented in References 4 and 5 were arrived at using conservative boending input parameters which encompassed the operating plants within the Westinghouse owners Group on RCS overpressurization, j1 Those generic algorithms concerning pressuee transients resulting from both mass and heat input to the RCS were uged with Plant-specific data to i arrive at Trojan-specific inlet fluid conditions. The original ar.alyses reported in References 2 and 3 have since been updated to reflect a n longer observed relief valve opening time. These analyses have shown that for liquid discharge, the range of expected PORY inlet fluid conditions is conscrvatively bounded in pressure by 15-670 psia, and in temperaturt by 1CO-500*F.
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t Trojr.n Nuciscr Plcnt Mr. S. A. Vergs Dockst 50-34a Attachment
. License NPF-1 December 31, 1985 Page 2 of 16 1
In the case of steam-liquid discharge, an analysis of the operating procedures, the OMS PORV resent pressures, and the applicable portions of Reference 1 have revealed that the range nf expected PORY inlet fluid conditions is conservatively bounded in pressure by 15-2575 psia, and in temperature by 100-572*F (liquid).
For steam discharge, an analysis of the operating procedures, the GMS PORY resent pressures, and the applicable portions of Reference 1 have revealed that the range of expected PORV inlet fluid cGnditions is conservatively bounded in presspre by 415-2375 psia, and in temperature by 448-672*F.
t
,' NRC Ouestion 2 .
In valve operability discussicas on cold oUcepressurination transients the submittal only idantifies conditions for water discharge. According to the Westinghouse Jatve inlet fluid conditions report, the PORVs are expected to operate over a range of steam, steam-water, and water conditions because of the potential presence of a steam buth)e in the pressurizer and w ter solid operations. To assure that the PORVs operate for all told overpressure events, discuse the range of fluid conditions erpected for the expected types of fluid discharge and identify the test i
data that dsmonstratee cperability for these cases. Since no Lcw pressure steam tests were performed for the PORVs, confirm that ths high pressure steam case for both opening and closing of the PORVs . . .
p [ demonstrates valve operability for the low pressure case).
PGE Response 2 The range of fluid conditions expected during a cold overpressure event are described in the above response to Question 1. Table 3-15 of the Trojan Plant-Specific submittai (PCE-1039) lists, and Section 3.3.2.2 of PGE-1039 describes in detail, the acceptability of the test data that demonstrates opcrability for the expected fluid conditions.
EPRI test conditions for the PORVs were thosen based on expected inlet fluid conditions, tests were limited but designed to confirm operability -
l over e Cull range of expected inlet conditions. Steam, stean-to-water and watse flow tests were conducted. Although steam tests were conducted i only at the higher pressures, Westinghouse has assured PCE that satisfactory operation would also result at the less sever 6 lower pressures. This can be seca by the successful low pressure, low tempetature water tests.
NRC Ouestion 3 Resulta from the EPRI tests on the Crosby safety val?es indicate that the test blowdowns axceeded the design value of 5% for both the "as installed" and " lowered" ring settings. The Trojan subntittal stated that i blowdowns in excess of the ASME specified design 5% blowdown was not I
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c Trojtn Nuciccr Pltnt Mr. S. A. Varge Dockst 50-344 Attachment License NPF-1 December 31, 1985 Page 3 of 16 considered significant from a nuclear safety standpoint. If the expected blowdowns for Trojan exceed 5%, the higher blowdown could cause a rise in pressurizer water level such that water may reach the safety valve inlet line and result in a steam-water flow situation. Also the pressure might be sufficiently decreased such that adequate cooling might not be achieved for decay heat removal. Provide additional discussion concerning these consequences and justification that nuclear safety is not inpaired.
PGE Response 3 PGE-2039 identified that Crosby HB-BP-86 (6M6) safety valve closure was abrupt with 4.8-12.7 percent blowdown. The Trojan safety valves are set to lift at 2485 psig. With a maximum 12.7 percent blowdown, the lowest pressure reached would then be 2169 psig. At this pressure, automatic mitigating and alarm functions that have initiated are: (a) full energization of pressurizer proportional heaters (2220 psig), (b) backup heaters (2210 psig), and (c) the pressurizer low pressure alarm (2210 psig). With pressurizer pressure at 2169 psig, there is still a large amount of margin until the pressurizer pressure low reactor trip setpoint (13b5 psig) or the saftty injectioa setpoint (1835 psig) is reached.
t Because a pressure transient resulting from 12.7 percent blowdown presents no challenge to Plant protection equipment, it is not considered significant from a nuclear safety standpoint.
PGE-1039 also identified valve tests that showed the Crosby safety valve, with rings set correctly, performed acceptably under the steam, steam-to-water, and water inlet fl;ild conditions that would result if the pressurizer fillzd solid during a maximum 12.7 percent blowdown.
NRC_ Question 4 The EPRI Inlet Fluid Justification Report suggested a method for demonstrating safety valve Jtability. This method compares the total inlet piping pressure drop for the in-Plant safety valves and piping to the applicable EPRI test safety valve piping combinations. The total inlet piping prescure drop is composeo of a frictional and acoustic wave i
component evaluated under steam conditions. The Trojan submittal did not provide pressure drop calculations but only compared piping lengths.
Provide the necessary pressure drops for the Trojan expected inlet conditions, flow capacity, ring settings, and inlet piping configuration. Make a comparison with the applicable EPRI test pressure
- drops to demonstrate valve stability.
, PGE Response 4 PCE-1039 referenced a report (Reference 6 of this letter) that provided Trojan-specifiq inint piping pressure drop calculations composed of both a frictional and acoustic wave component ovaluated under steam i conditions. The necessary inlet piping pressure drops were Jisted in l
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l Terjcn Nucloce Pltnt Mr. S. A. Varg3 Docket 50-344 Attachment ;
License NPF-1 December 31, 1985 Page 4 of 16 Table 3-10 of PGE-1039. The pressure drop calculations are presented below for your reference:
Trojan Plant-Specific Pressure Drop Calculations
- 1. Transient Flow Pressure Difference (app) Calculation The flow pressure difference due to pipe friction and fittings is given by:
- If T s 2L/a, IL .2 APF = (K + D ) (CM) 2gepA
- If T 1 2L/a, fL app = (K + D ) (CM)
.2'lk\2 taTi 2gepA K = summation of expansion and contraction loss coefficients corrected if required to correspond to the inlet piping flow area. (NOTE: The contraction from the pressurizer to the inlet pipe can be assumed to be smooth and, therefore, the loss coefficient can be assumed to be zero.) (dimensionless) f = friction factor (dimensionless).
L = equivalent length of a resistance to flow, in pipe diameters.
D R = rated valve flow rate for steam (lb/sec).
ge = gravitational constant (32.2 lb-ft/lb-sec 2),
p = stean density at nominal valve set pressure (lb/f t 3)
A = inlet piping flow area (ft 2),
a = steam sonic velocity (ft/sec) - use 1100 ft/see for all calculations.
L = inlet piping length (from the pressurizer inside diameter to the interface between the inlet pipe flange and the valve inlet flange) (ft).
T = valve opening or closing time for steam inlet conditions (sec).
Trejan Nuc122r Plcnt Mr. S A Varge Docket 50-344 Attachment License NPF-1 December 31, 1985 Page 5 of 16 C = flow rate constant for valve opening or closing.
AP = Differential pressure (psi).
- 2. Transient Acoustic Wave Amplitude (APAW) Calculation The acoustic wave amplitude is calculated based on information in Reference 7. There are two situations to consider:
- If T 1 2L/a, a(CR) + (CR)
APAW " 2 sAc 2gepA
- If T > 2L/a, 2L 2 2L (CR) + (CR)2 iT; APAW " 2 s cAT 2gepA All parameters are defined in Section 1 above.
- 3. Plant-Specific Transient Pressure Difference Calculation The Plant-specific transient pressure difference associated with valve opening or closing is equal to the sum of the flow pressure difference (app) and the acoustic wave amplitude (APAW) as determined above.
- 4. Plant-Specific Steady-State Flow Pressure Difference (APF )
Calculation The steady-state flow pressure difference associated with valve opening or closing is given by:
ik . 2 (K + D ) (CM) app = 2gepA 2 All parameters are defined in Section 1 above. Note that the values of the flow rate constant, C, are different for valve opening.and closing.
- a. Valve Opening .
The opening time is:
Top = .010 sec.
m 4
Trajan Nuclear Plant Mr. S. A. Vargs Docket 50-344 Attachment License NPF-1 December 31, 1985 Page 6 of 16 Also, 2L , 2 x 8.8 ft . 016 sec, a 1100 ft/sec.
Therefore, Top < 2L/a (1) Transient Flow Pressure Difference Since Top < 2L/a, the flow pressure difference is:
APy = (K + f1/D)(CR) 2gepA Where, R = 420.000 lb/hr = 116.7 lb/sec. 3600 sec./hr C = 1.11 K = 0 - see Item 1 above f = .015 L = 8.8 + 1x(16) + 3x(30) = 126.4 D .432 p = 7.65 lb/ft 3 A = 0.147 ft2 The flow pressure difference is:
APy = [0+(.015)(126.4)][(1.11)(116.7)]2 64.4 x 7.65 x .1472 x 144 APy = 20.8 psi (2) Transient Acoustic Wave Amplitude Since Top < 2L/a, the acoustic wave amplitude ist APAW = a(CR) + (CR) geA 2g cpA
= (1100)(1.11 x 116.7) + (1.11 x 116.7) 32.2 x .147 x 144 64.4 x 7.65 x .147 x 144
r-Tr2jan NuclOcr Plcnt Mr. S. A. V rgs Docket 50-344 Attachment License NPF-1 December 31, 1985 Page 7 of 16 APAW = 219.9 psi (3) Plant-Specific Transient Pressure Difference The Plant-specific pressure difference for valve opening is, AP = APy + APAW
20.8 + 219.9
AP = 240.8 psi (4) Plant Specific Pressure Difference The steady-state flow pressure difference for valve opening is, fL 2 APy = (K + D ) (CR) 2SepA APy = 20.8 psi (5) Plant-Specific Pressure Difference for Plant Versus Test Evaluation (opening)
Based on the above, the controlling pressure difference is the transient pressure difference is 240.8 psi.
- b. Valve closing The closing time is, TCL = .016 C = .69 Also, 2L/a = .016 sec.
TCL s .016 sec. 1 2L/a (1) Transient Flow Pressure Difference is, fL APF = (K + D") (CR) 2gepA APy = (O. + .015 x 126.4) x (.69 x 116.7)2 64.4 x 7.65 x .1472 x 144 APy = 8.02 psi
_ _ . . = - - . _ .
Trajcn Nucleer Plcnt Mr. S. A. Vzrgs Docket 50-344 Attachment License NPF-1 December 31, 1985 Page 8 of 16 (2) Transient Acoustic Wave Amplitude Since TCL i 2L/a, the acoustic wave amplitude is, APAW = a(d) + (d) geA 2g pA
= (1100)(.69 x 116.7) + (.69 x 116.7) 2 32.2 x .147 x 144 64.4 x 7.65 x .147 x 144 APAW = 134.2 psi (3) Plant-Specific Transient Pressure Difference The Plant-specific pressure difference for valve closing is, A = app + APAW = 8.02 + 134.2 = 142 psi (4) Plant-Specific Steady-State Flow Pressure Difference The steady-state flow pressure difference for valve closing is the same as for valve opening (20.8 psi).
(5) Plant-Specific Pressure Difference for Plant Versus Test Evaluation (Closing)
Based on the above, the controlling pressure difference for Trojan is the transient pressure difference, ie, 142 psi.
NRC Ouestion 5 The Westinghouse inlet fluid conditions report stated that liquid flow could exist through the PORV for the FSAR feedline break event and the extended high pressure injection event. Liquid PORY flow is also predicted for the cold overpressurization event. These same flow conditions will also exist for the Block Valve. The EPRI/ Marshal Block Valve Report did not test the block valves with fluid media other than -
steam. The Westinghouse Gate Valve Closure Testing Program did include tests with water; however..the information presented in the report did not provide specific test results. Since it is conceivable that the ENOV could be expected to operate with fluid flows, discuss gMOV block valve
- . operability with expected liquid flow conditions and provide specific test data.
t PGE Response 5 In a June 1, 1982 Westinghouse letter from R. C. Youngdahl to Mr. H.
Denton, several block valve test submittals were made which included an
F.
Trejan Nuclair Plcnt Mr. S. A. Yzrgs Docket 50-344 Attachment License NPF-1 December 31, 1985 Page 9 of 16 explanation as to why block valve tests beyond the Marshall tests were not considered necessary, as well as an EPRI summary report covering Westinghouse gate valve closure testing. The Westinghouse report, transmitted to the NRC by the Youngdahl submittal, also includes a section on friction testing of stellited seating parts.
Friction testing done by Westinghouse on stellite test specimens (Note:
the Velan valve also has stellite seats) indicates that over the initial 200 cycles of testing, water test specimen friction factors increased from as low as 0.12 until a level of 0.4 to 0.75 is reached. With 550*F steam, the friction factor starts in the 0.5 to 0.6 range (higher than the water tests) and drops approximately 0.35 over the 200 cycle range.
Considering the 21 test cycles completed at Marshall Steam Station, and in view of the above frictional data, the thrust required to cycle the valve during the steam tests would be similar to that if the test medium were water.
NRC Question 6 Bending moments are induced on the safety valves and PORVs during the time they are required to operate because of discharge loads and thermal expansion on the pressurizer tank and inlet piping. Make a comparison between the predicted plant moments with the moments applied to the tested valves to demonstrate that the operability of the valves will not be impaired.
PGE Response 6 Pipe bending moments induced on safety valves and PORVs during their valve actuation are summarized by Tables 1 and 2. The analyzed moments are compared to the maximum tested moments for both the three Crosby HB-BP-86 6M6 safety valves, and the two Copes-Vulcan Model D-100-160 relief valves (316 w/ stellite plug and 17-4PH cage). Relief valve accelerations, as calculated at the valve assembly center-of-gravity, are compared to the tested accelerations. Tested moments for the relief valves were obtained for the same Copes-Vulcan valve body (as installed) with a 17-4PH plug and cage.
Maximum analytical moments and accelerations are less than those for i
valves successfully tested (without damage affecting valve performance).
! This comparison ensures the operability of the installed valves.
! NRC Question 7 l The Westinghouse Valve Inlet Fluid Conditions Report states that liquid discharge could be expected through the safety valves for both the feedline break and extended high pressure injection events. The RPRI 6M6
! test safety valve experienced some chatter and flutter while discharging liquid at certain ring settings. Testing was terminated af ter observing chattering to minimize valve damage. Inspection revealed some valve I
l s
Trojan Nucicar Plent Mr. S. A. Vress Docket 50-344 Attachment License NPF-1 December 31, 1985 Page 10 of 16 damage which was presumably caused by the valve chatter and flutter.
Liquid discharge for Trojan may conceivably occur for longer periods of time than the EPRI testing. Thus, longer periods of valve chattering may cause severe valve damage. Discuss the implications this may have on operability and reliability of the Trojan safety valves. Identify any actions that will be taken to inspect for valve damage and assure reliable operability following safety valve lift events.
PCE Response 7 The Trojan FSAR states that liquid discharge will be expected for the feedline break. It is agreed that at certain ring settings (not those associated with normal Trojan operation), the crosby 6M6 safety valve !
experienced some chatter and flutter while discharging liquid during testing. However, at normal Trojan ring settings and fluid inlet conditions expected during a main feedline break or spurious safety injection at power, the safety valves had stable performance and operated acceptably. In addressing observed valve performance, one must differentiate between the valves and fluid conditions tested and the actual valves and actual fluid conditions for the specific plant. The EPRI inlet piping arrangement, flow and acoustic pressure drops, and inlet fluid conditions bound the same plant-specific parameters for Troj an. Valve performance observed during the EPRI tests, therefore, reflects worst case performance as compared to results that would be observed had the testing been conducted using actual plant-specific piping arrangements and fluid conditions. The main feedline break and the spurious safety injection at power events were both bounded by EPRI Test 931a. This was a low flow steam-to-water transition test with a filled loop seal and high back pressure. The nominal temperature of the water in the loop seal was 100*F at the valve inlet. The valve actuated twice during this test with the steam-to-water transition occurring 14 seconds into the first actuation. Therefore, the Trojan safety valves are expected to operate satisfactorily during a main feedline break or spurious safety injection at power. Because of the classification of a main feedline break as ANS Condition IV, this severe type of accident is not expected to take place, but if it does, evaluations will be performed I as to what actions will be necessary to inspect for valve damage and I assure reliable operation following safety valve lift events. l NRC Ouestion 8 NUREG-0737 Item II.D.1 requires that the plant-specific PORV control l circuitry be qualified for design-basis transients and accidents. Please provide information which demonstrates that this requirement has been ,
fulfilled. 1 l
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Trejrn Nu210cr Plcnt Mr. S. A. Vargs Docket 50-344 Attachment License NPF-1 December 31, 1985 -
Page 11 of 16 PCE Response 8 We have verified that the PORV control circuitry is indeed qualified for design-basis transients and accidents as follows:
- a. The solenoid valves that control the air to the power-operated relief valves are environmentally qualified (EQ).
- b. The power supplies for the solenoid valves are Class 1E from the safety-related 125-V de buses,
- c. The cables to the solenoid valves are in channelized raceways.
- e. All of the electronic control modules are located in the seismically-qualified Hagan cabinets in the control room.
- f. All of the applicable Hagan modules are powered from preferred 120-V ac buses.
L g. The manual selector switches and manual / automatic control stations are located in the seismically-qualified main control panels in the control room. The selector switches are electrically separated by the enclosure of one switch in a metal box, and the enclosure of cable in flexible conduit.
NRC Ouestion 9 The Trojan plant safety valves are Crosby 6M6 and were tested by EPRI.
EPRI testing of the 6M6 was performed at various ring settings. The submittal did not provide the present ring settings but stated the rings were set at the factory recommended settings. If the plant current ring settings were not used in the EPRI tests, the results may not be directly applicable to the Trojan safety valves. Identify the Trojan safety valve ring settings. If the plant specific ring settings were not tested by EPRI, explain how the expected values for flow capacity, blowdown, and
- the resulting back pressure corresponding to the plant-specific ring settings were extrapolated or calculated from the EPRI test data.
Identify these values so determined and evaluate the effects of these values on the behavior of the safety valves.
PGE Response 9 In the EPRI tests, positions of the two adjusting rings, termed upper (guide ring) and lower (nozzle ring), are given in notches relative to the level position. The level position is the number of notches relative to the upper limit of ring travel for which the bottom of the upper ring is flush with the bottom of the disc ring. Plant (fleid) settings for
. the upper ring are provided by Crosby in notches relative to the s
--,-_r--- - . - , , - - , . - - - , - - - - - - , - - - - - - - - . - ,
Trajcn Nuclect Pltnt Mr. S. A. Vargs Docket 50-344 Attachment License NPF-1 December 31, 1985 Page 12 of 16
" highest-locked position", which is the upper limit of ring travel. The as-tested and field settings for the upper ring are, therefore, not directly comparable. This relationship is being researched and will be reported to the NRC as soon as it is obtained from Crosby. For the lower ring, the " highest-locked position" coincides with the level position; therefore, both Plant and test settings are directly comparable.
Based on available information, the Trojan pressurizer safety valves have an upper ring setting of -285 and a lower ring setting of -28. The typical Plant settings used in the EPRI tests were -71 and -77 for the upper ring setting, and -18 for the lower ring setting. The difference in upper ring settings is believed to be due to the difference in where the ring setting is measured from (as discussed above). The difference in lower ring settings is believed to be insignificant. Lowering the lower ring setting (making it more negative) will result in a lengthening of the valve opening time, which was observed in the EPRI testing to have a positive effect in preventing valve chatter.
It should be noted that the Trojan safety valve ring settings remain those that were set at the factory. Ring positions, based on manufacturer's recommendations and representative of those utilized in typical Plant installations, provided the primary input to the selection of ring settings actually used in the EPRI testing.
NRC Ouestion 10
, The submittal states that a hydraulic analysis of the safety / relief valve piping system has been conducted. To allow for a more complete evaluation of the methods used and the results obtained from the
, thermal-hydraulic analysis, provide additional discussion on the thermal-hydraulic analysis that contain at least the following information:
- a. An explanation of the method used to treat valve resistances in the analysis. Report the valve flow rates that correspond to the resistances used. Because the ASME Code requires derating of the safety valves to 90 percent of actual flow capacity, the safety valve analysis should be based on flows equal to 111 percent of the valve flow rating, unless another flow rate can be justified. Provide information explaining how decating of the safety valves was handled
.and describe methods used to establish flow rates for the safety valves and PORVs in the analysis.
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- b. A discussion of the sequence of opening of the safety valves that was used to produce the worst case loading conditions.
- c. Copies of the EDS Nuclear, Inc. thermal analysis reports.
I
4-Trojan Nucl0cr Plcnt i Mr. S S A. Vargs Docket 50-344 ; k Attachment.
License NPF-1 N December 31', 1985 I
Page 13 of 6 i.
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PCE Response 10 0 7
., 9
.a. The valve resistance is selected to ensure choked flow conditions at the valvo and the desired valve flow rate of 485.100 lbm/hr. The rated flow rate for the Crosby 6M6 is 420,000 lbm/he;. consequently, the value used is 15 percent higher and more than compensates for the ASME derat'ng. ,
The PORVs are Copes-Vulcan Model D-100-160. The maximum flow rate based on the specification is 210,000 lbm/hr. The EPRI tested flow rates of 255,600 lbm/hr (not derated) were used.
+ !
Since choked flow conditions exist at the valves, appropriate resistances and flow area were selected to ensure that the desired flow rates werelattained.
., , t
- b. Sequential actuation of the relief and the safety valves by design setpoint (as oppcsed to simultaneous valve actuation) represents a realistic transient of the cold loop seal followed by steam discharge. -This was used to generate the forcing functions,
- c. For IMPELL (EDS Nuclear) thermal analysis report see IMPELL Stress Analysis Report No. 01-0300-1291 (Reference 8). This report is included as part of,this attachment.
NRC Ouestion il The submittal states that a structural analysis of the safety and relief valve piping system has been conducted. To allow for a more. complete evaluation of the methods used and results obtained from the structural analysis, please provide reports containing at least the following ,
information: <
- a. The REFORC 2 program was' identified as the program used in the analysis. How was the program verified? '
- b. A description of methods used to model supports, the pressurizer and relief tank connections, and the safety valve bonnet assemblies and l- PORV actuator.' ,-
4
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- c. An identification of the load' combinations performed in the analysis together with the allowable: stress limits. Differentiate between load combinations used in the piping upstream and downstream of the valve. Explain the'mathematicaf methods used to perform the load i
combinations, and identify the. governing codes and standards used to ,
[ determine piping and support adequacy.
- d. A sketch of the structur 1 model showing lumped mass locations and i pipe sizes.
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f; Tr@jtn Nucisar Pirnt Mr. S. A. Vargs Docket 50-344 Attachment License NPF-1 December 31, 1985 Page 14 of 16 1 l
- e. Copies of the EDS Nuclear, Inc. structural analysis report.
I PGB Response 11 j
- a. REFORC 2 is a post-processor program developed by IMPELL Corporation ,
for the Electric Power Research Institute in support of the' '
safety / relief system requalification effort.
1 The program was verified in accordance with the IMPELL QA Program using hand calculations for subcooled liquid, saturated liquid with flashing and two-phase flow, while hand calculations and Moody (Reference 13, pp 219-260) were used for superheated steam conditions,
- b. Large bore pipe, pressurizer and relief tank support stiffnesses are calculated and input into the pipe stress computer. analyses.
Stick models of both the pressurizer and pressure relief tank are included in the analyses as shown by the IMPRLL Large Bore Stress Report (Reference 8. Appendix A math model). The pressurizer is modeled with equivalent mass, cross-sectional diameter, wall .
thickness and material. The pressurizer relief tank is modeled with equivalent cross-sectional diameter, wall thickness, and material in order to accurately account for significant dynamic and thermal expansion effects of the equipment on the pipe.
The PORY and safety valves are modeled with the entire valve / operator assembly mass lumped at the center-of-gravity.of the entire assembly. Mass of valve attached flanges are lumped at their center-of-gravity. The safety valve bonnet assembly and PORY actuators are modaled with stiffnesses considerably higher than attached pipe. This was reviewed to ensure that possible dynamic excitation with the pipe was not significant.
- c. Load combinations with allowable stress limits are provided in the IMPELL Large and Small Bore Stress Reports (References 8 and 10).
Copies of both of these references are attached. These reports identify the governing codes and standards, and explain the mathematical methods used in load combinations for piping component qualifications. Qualification of trunnion and anchor to pipe attachment interfaces are included with the piping qualifications.
- d. Sketches of the piping structural models with pipe sizes are provided by the IMPELL Large and S tall Bore Stress Reports (References 8 and 10, as Appendix A "Matheraatical Model").
Masses are analytically lumped at the numbered joints shown by the Mathematical Models for valves, flanges, and analytically decoupled large bore branch pipe connections. Distributed pipe masses are analytically lumped at numbered joints for dynamic analyses. When necessary, additional mass joints are placed automatically by the computer program SUPERPIPE (Reference 11) between joints.
Trojcn Nucl00r Pirnt Mr. S. A Vargs Docket 50-344 Attachment License NPF-1 December 31, 1985 Page 15 of 16
- e. Ccpies of the IMPELL (EDS Nuclear) structural analysis reports (References 8 and 10) are included as part of this attachment.
NRC Ouestion 12 According to results of EPRI tests, high frequency pressure oscillations of 170-260 Hz typically occur in the piping upstream of the safety valve while loop seal water passes through the valve. An evaluation of this phenomenon is documented in the Westinghouse report WCAP-10105 and states that the acoustic pressures occurring ptior to and during safety valve discharge are below the maximum permissible pressure. The study discussed in the Westinghouse report determined the maximum permissible pressure for the inlet piping and established the maximum allowable bending moments for Level C Service condition in the inlet piping based on the maximum transient pressure measured or calculated. While the internal pressures are lower than the maximum permissible pressures, the pressure oscillations could potentially excite high frequency vibration modes in he piping, creating bending moments in the inlet piping that should be combined with mcments from other appropriate mechanical loads.
Provide one of the following: (1) a comparison of the expected leak pressures and bending moments with the allowable values reported in the WCAP report, or (2) justification for other alternate allowable pressure and bending moments with a similar comparison with peak pressures and moments induced in the plant piping.
PGE Response 12 A review was performed of the dynamic modes of the piping upstream of the safety valves. This review verified that none of the significant modes of the upstream piping falls within the frequency range of the pressure oscillations documented in the WCAP report.
To provide further documentation, an IMPELL calculation was generated (Reference 12). This calculation, using very conservative calculation techniques, provides sufficient justification to indicate that Service
,_ Level C stresses will remain below the Service Level C limit when
( stresses due to high-pressure oscillations are included.
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Trajtn Nuclear Plcat Mr. S. A. Vcrgs l Docket 50-344 Attachment License NPF-1 )
December 31, 1985 ;
Page 16 of 16 REFERENCES
- 1. Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valves in Westinzhouse-Designed Plants. EPRI NP-2296-LD, PWR Safety and Relief Valve Test Program (March 1982).
- 2. PGE (C. Goodwin) to NRC (A. Schwencer) letter dated July 21, 1977.
- 3. PGE (C. Goodwin) to NRC (A. Schwencer) letter dated February 28, 1978.
- 4. Pressure Mitimating Systems Transient Analysis Results. Prepared by Westinghouse Electric Corporation for the Westinghouse Owners Group on Reacter Coolant System Overpressurization, July 1977.
S. Supplement to the July 1977 Report. Pressure Mitigating Systems Transient Analysis Results Prepared by Westinghouse Electric Corporation for the Westinghouse Owners Group on Reactor Coolant System Overpressurization, September 1977.
- 6. PWR Safety and Relief Velve Adequacy Report for Portland General Electric Company Trojan Nuclear Plant, Prepared by R. M. Grayson of the Westinghouse Electric Corporation December 1982.
.7. Waterhammer Analysis, John Parmakian, Dover Publications, Inc., 1963.
- 8. IMPELL Report No. 01-0300-1291, Trojan Nuclear Power Plant Modification and Stress Analysis Report for the Pressurizer Safety and Relief Valve Large-Bore Piping System, Revision 0, dated April 1984.
- 9. EPRI Special Report No. NP-2628-SR, "EPRI PWR Safety and Relief Valve Test Program, Safety and Relief Valve Test Report", dated December 1982.
- 10. IMPELL Report No. 01-0300-1292, Trojan Nuclear Power Plant Modificaiton and Stress Analysis Report for Pressurizer Safety and Relief Valve Small-Bore Piping System, Revision 0, dated June 1984.
- 11. IMPELL Computer Program SUPERPIPE, Version 16A dated October 1, 1983.
- 12. IMPELL Calculation RC-08 Job No. 0300-027, "NUREG-0737 Item II.D.1 NRC Question Response", Revision O.
l 13. Moody, F.J., " Fluid Reaction and Impingement Loads", Nuclear Power Plants.
GVB/kal 1521G
P CROSBY KB-BP-86-6M6 DISCHARGE INDUCED MOMENT COMPARISON TABLE 1 Crosby HB-BP-86-6M6 Maximum Safety Relief Valve Analyzed Moment (2) Tested Moment (3)
Trojan Valve No. Joint No.(1) (Ft-Lbs) (Ft-Lbs)
PSV 8310A 89 21,258 24,896 PSV 8010B 54 22,633 24,896 PSV 8010C 18 24,686 24,896 NOTES:
- 1. Joint numbers are per Reference 8 Appendix A math model at valve discharge end.
- 2. Analyzed moment = deadweight absolutely summed with envelope of possible thermal expansion cases (all valves closed and all valves opened), and safety valve opening thrust moments. Moments are per Reference 8. Appendix C.
- 3. Tested moment is per Reference 9 (Pgs. 3-69 and 3-71).
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l COPES-VULCAN DISCHARGE-INDUCED MONENT AND ACCELERATION COMPARISON TABLE 2 Copes-Vulcan Relief Valve Model Maximum Maximum D-100-160 (316 W/ Stellite Analyzed 2 Tested Analyzed Tested Plus'and 17-4PH Cane) Moment Moment 3 Acceleration 4 Acceleration 5 1 Trojan Valve No. Joint No.1 (Ft-Lbs) (Ft-Lbs) (G) (G)
PCV-455A- 163 1778 3583 - -
V5A - -
3.1 12.6 PCV-456 148 3311 3583 - -
V6A - -
4.4 12.6 NOTES:
- 1. Joint numbers are per Reference 8, Appendix A, at valve discharge ends for
, moments, and at valve assembly center of gravity for accelerations.
- 2. Analyzed moment = deadweight absolutely sunned with envelope of possible thermal expansion cases (all valves closed, relief valve 455A open, relief valve.456 open, both relief valves opened, and'all safety and relief valves open), and relief valve ~ opening thrust moments. Moments are per
< Reference 8, Appendix C.
j 3. Tested moment is per Reference 9 (Pages 4-70).
- 4. Analyzed accolerations during rolief valve discharge are per Reference 8 Appendix C.
- 5. Tested accelerations are per Reference 9 (Pages 4-62).
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