ML18019B065
| ML18019B065 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 07/03/1986 |
| From: | Zimmerman S CAROLINA POWER & LIGHT CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM NLS-86-250, NUDOCS 8607100238 | |
| Download: ML18019B065 (33) | |
Text
REGULA Y INFORl'fATIQN DISTR IBUTI YSTEN ( R IDS)
ACCESSION NBR: 8607100238 DQC. DATE: 86/07/03 NOTARIZED: NO DOCKET FACIL: 50-400 Shearon Harris Nuclear Power Planti Unit ii Carolina 05000400 AUTH. NANE AUTHOR AFFILIATION ZINNERNANi S. R. Carolina Power 5 Light Co.
RECIP. NANE RECIPIENT AFFILIATION DENTON, H. R. Of f ice of Nuclear Reactor Regulationi Director (p ost 851125
SUBJECT:
'Forwards addi info requested in NRC 860512 ltr re testing of safety 8r relief valves. Info submitted to allow close out of SER Confirmatory Item 6 5 NUREG-0737'tem II. D. 1.
DISTRIBUTION CODE: A046D COPIES RECEIVED: LTR ENCL SI ZE:
TITLE: QR Submittal: Tl'fl Action Plan Rgmt NUREG-0737 Cc NUREG-0660 NOTES: Application for permit renewal filed. 05000400 RECIPIENT COPIES RECIP I ENT COPIES ID CODE/NANE LTTR ENCL ID CODE/NANE LTTR ENCL PWR-A ADTS PWR-A EB 1 1 PWR-A EICSB 2 2 PWR-A FOB PWR-* PD2 LA 0 PWR-A PD2 PD 01 5 5 BUCKLEY B 1 1 PWR-A PSB PWR-A RSB 1 1 INTERNAL: ADN/LFNB 1 0 ELD/HDS1 0 IE/DEPER DIR 33 1 1 IE/DEPER/EPB 3 3 NRR BWR ADTS 1 NRR PAULSON W. 1 NRR PWR-A ADTS 1. 1 NRR PWR-B ADTS 1 NRR/DHF 1 NRR/DSRO ESPRIT 1 1 1 RGN2 EXTERNAL LPDR 03 NRC PDR 02 NSIC 05 TOTAL NUNBER OF COPIES REQUIRED: LTTR 31 ENCL 28
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CNK Carolina Power & Light Company SERIAL: NLS-86-250 JUL 0 S 1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NO. 1 - DOCKET NO.50-000 TESTING OF RELIEF AND SAFETY VALVES
Dear Mr. Denton:
Carolina Power R Light Company (CPRL) hereby submits additional information concerning testing of-safety and.relief. valves for;.the Shearon Harris Nuclear Power Plant. The attached information;is submitted in response to an NRC request for additional information transmitted by letter dated May 12, 1986. This information is being submitted to allow closeout of Safety Evaluation Report Confirmatory Item No. 6 and NUREG-0737 Item II.D.1.
If you have any questions on this subject or require additional information, please contact me.
Yours very truly, S. Z'erman Manager Nuclear Licensing Section 3HE/crs (00003DK)
Attachment cc: Mr. B. C. Buckley (NRC)
Mr. A. S. Masciantonio (NRC-PAEB)
Mr. G. F. Maxwell (NRC-SHNPP)
Dr. 3. Nelson Grace (NRC-RII)
Wake County Public Library 8b07100238 8b0703 PDR ADOCK 05000400 E PDR
/IIov4 411 Fayettevilte Street o P. O. Box 1551 ~ Raleigh, N. C. 27602
'(r
QUESTION 1: Single Failures The submittal (Reference 1) did not include a discussion on the consideration of single failures after the initial events that challenge the PORVs and/or safety valves. NUREG 0737 requires of single failures that produce maximum loads on the 'election PORVs and safety valves. Provide a discussion describing how the single failure considerations are met.
RESPONSE
NUREG 0737 requires selection of single failures that produce maximum loads after the initial events that challenge the Power Operated Relief Valves (PORV's) and/or safety valves (SRV's).
The case of three PORV's or three safety relief valves opening simultaneously produces higher loads than the failure of one PORV and/or one relief valve to open. Mhen one of the valves (PORV or safety relief) does not open, the loads immediately downstream of the other two valves will be identical to the loads corresponding to the case of all three valves opening.
The loads on the piping elsewhere are lower because of the reduced overall flow rate. The case of the three safety valves opening while the PORV's remain closed produces higher loads downstream of the SRVs than if any or all of the PORV's opened and the SRVs open against a resultant higher downstream pressure.
For stuck-open relief or safety valves, steam, sub-cooled water or two phase flow can be expected. However, the pressure upstream of the valve is below the valve setpoint pressure, the downstream piping is already pressurized and the fluid is at high temperature. Therefore, the loads are lower than those occurring during the acceleration of the cold loop seal in the beginning of the transient.
IBMD-PROC01-OSR Question/Answer
In discussion of the PORV inlet condition for cold overpressure transients, the submittal only identified the temperature and pressure as approximately 100 to 350'F, and 400 to 2,335 psig.
However, the PORV is expected to operate over a wide range of temperature and pressure conditions during steam and water discharge for cold overpressure events. To assure that the PORVs operate under all cold overpressure transients, discuss the range of fluid conditions expected for the various types of discharge (including the nitrogen bubble case) and identify the EPRI tests that demonstrate operability over the entire range of conditions. Confirm that the high pressure steam tests demonstrate valve operability for the low pressure steam case.
RESPONSE
The two PORV's which provide protection against low temperature overpressurization events are 3" Copes-Vulcan stainless steel valves with stellite plug. This valve was tested by EPRI in response to NUREG 0737 Item II.D.1.A requirements. The EPRI program consisted of a wide variety of tests enveloping fluid transients with different valve inlet conditions.
The program results are provided in EPRI's "Safety and Relief Valve Test Report" (Reference 5). This report demonstrated that the valve fully opened and closed on demand at the Marshall Steam Station and during Phase III of the Wyle Test Program. Table "A" compares the expected plant conditions at the PORV inlet with the tested conditions in the above mentioned EPRI report. Based on this comparison, it is concluded that the tested conditions are representative of expected conditions for cold overpressurization events at Shearon Harris Nuclear Power Plant.
The high pressure steam tests demonstrate valve operability for the low pressure steam case. For higher pressures, the flow rates and the energy releases are higher, consequently, the acoustic wave amplitudes and flow pressure differences in the inlet piping are also higher.
IB1fD-PROC01-OSR Question/Answer
TABLE A PLANT CONDITIONS TESTED CONDITIONS Upper Peak PORV Pressurizer Pressurizer Inle t Inlet Setpoint Pressure Temperature Fluid Test 'ressure Temp. Fluid (PSIG) (PSIA) ( F) State No. (PSIA) ('F) State 400 492 85-300 Water 74-CV-316-5W 675 105 73-CV-316-4W 675 442 Water 410 501 325 Water 73-CV-316-4W 675 442 Water 410 505 450 Water- 76-CV-316-2W 2535 647 Water Steam 77-CV-316-7S/W 2532 670 Transi-tion Steam 435 550 500 Steam 71-CV-316-1S 2715 682 Steam 450 581 535 Steam 71-CV-316-1S 2715 682 Steam 2400 2350 550 Steam 71-CV-316-1S 2715 682 Steam NOTE: 1. The SHNPP PORVs are not expected to operate under a nitrogen bubble and hence the nitrogen bubble case has not been evaluated.
IBMD-PROC01-OSR Question/Answer
Plots of the predicted loop seal temperature distribution upstream of the safety valve and PORV are presented in the discharge piping analysis report prepared by EBASCO (attachment to reference 1). Since the analysis of the inlet piping was performed by a different contractor and was not provided in the submittal, it is not known whether the temperature distribution used by EBASCO in the discharge piping analysis was also used to analyze the inlet piping as well. Confirm that the loop seal temperatures presented in the discharge piping analysis report are representative of the temperature distributions in the safety valve and PORV loop seals and that the same temperature distribution was used in both the inlet and discharge piping analysis. State whether the above temperature data were verified by field measurements at the Shearon Harris plant and provide a comparison of the field measurements with the predicted values.
RESPONSE
The piping analysis for both the safety valve's and PORV's inlet and discharge lines were performed using an average loop seal temperature of 209'F. This value for temperature was obtained from testing done during the SHNPP Hot Functional Test (HFT). A more detailed discussion of that testing can be found in the June 6, 1986 submittal number NLS-86-197. Based on this as measured temperature, the piping stress analysis assumed no flashing of the loop seal volumes across the valves. This is conservative based on HFT data that demonstrates that some percentage of the loop seal volume is hot enough to flash, hence reducing the actual loads to the piping.
IBMD-PROC01-OSR Question/Answer
The submittal did not identify the ring settings for the safety valves. Give the ring setting values for the three safety valves. If the ring settings are given in terms of the field positions (i.e., referenced to the locked position) in your response, indicate which ring setting used in the EPRI tests is applicable to the in-plant ring settings and explain why.
RESPONSE
All three of the SRVs have lower ring settings of -18. This position is the same as level position and is directly in correlation to the EPRI settings. The upper ring settings are
-250 for two of the valves and -235 for one SRV. These settings are relative to the locked position. The upper ring settings noted in the EPRI tests are relative to the level position and can be converted as such. Valve 1RC-R528SN-1 has a ring setting of - 104, Valve 1RC-R529SN-1 has a ring setting of -81 and Valve 1RC-R530SN-1 has a setting of -91. EPRI evaluated the Crosby 6M6 Valves over a range from -44 to -186 which envelopes the SHNPP ring settings.
IBMD-PROC01-OS4 Question/Answer
In one of the EPRI steam tests on the Crosby 6M6 safety valve (Test No. 1419), the valve chattered on closing. The test was terminated after the valve was manually opened to stop the chatter. Compare the ring setting and loop seal temperature of the test valve with that of the Shearon Harris safety valves and determine whether Test 1419 is applicable to the evaluation of the operability of the in-plant valves. If this test is applicable, explain why the same closing problem experienced by the test valve will not happen to the Shearon Harris valves.
Otherwise, state what measures will be taken to ensure positive valve closure and prevent disk damage, if similar chattering should occur to the in-plant safety valves.
RESPONSE
Ring settings for the Shearon Harris safety valves are provided .
in the response to question 4. Loop seal temperatures have been previously provided most recently in CP&L's June 6, 1986 submittal, number NLS-86-197. For information, the temperatures in the loop seals average 209'F.
Test 1419 was one of several loop seal tests done with ring settings established by the same methods as the SHNPP ring settings. The other tests are numbers 1406, 1415, and 929.
For these four tests, the fluid temperature at the valve inlet ranged from 90'F to 350'F.
Tests 1415 and 1419 were nominal 350'F loop seal tests, while 1406 and 929 were nominal 100'F tests. In three of these four, the valve operated satisfactorily. There were nine additional loop seal tests done on 6M6 valves with different ring settings. In eight of these, the valve operated satisfactorily, the remaining test having the same results as 1419.
Taken together, these tests present a picture of valve operation that shows satisfactory performance: the valve does not leak excessively, it lifts to relieve pressure, and recloses. An anomaly occurred in test 1419 after the valve had closed.
Re-opening after closure is thought to be a function of the inlet pipe pressure oscillations> valve response time and inlet pipe length. The valve closing sets up oscillations that are reflected back to it. If the valve responds quickly, and the oscillations are not, damped out, the valve may begin to open in response to these oscillations and close to create further oscillations. Test 1419 is an isolated example of this.
IBMD-PROCOl-OSR Question/Answer
RESPONSE: (continued)
As the EPRI tests conditions were to envelope many plants, and the inlet pipe configuration and pressure drops for these tests envelope those at SHNPP, then safety valve operation at SHNPP can be expected to be at least as good as that of the applicable EPRI tests, which were generally satisfactory.
IBMD-PROC01-OSR Question/Answer
The submittal presented a summary of the EPRI test data on the Crosby 6M6 safety valve which indicated that the blowdown values for steam exceeded the 5% value given in the valve specifications by a wide margin. The higher blowdowns 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 decreased to such an extent that adequate cooling might not be achieved for decay heat removal. Discuss these consequences of high blowdowns if similar high blowdowns are expected for the Shearon Harris safety valves. Provide analysis results to demonstrate that the increased blowdown is not expected to have an adverse effect on plant safety, or submit a copy'of the Westinghouse report which presents their study, on behalf of the Westinghouse Owners'roup, on the consequences of the increased blowdown in Westinghouse designed PWRs, if the analysis and conclusion contained in that report are considered applicable to the Shearon Harris plant.
RESPONSE
The impact on plant safety of excessive pressurizer safety valve blowdowns (up to 14X) has been evaluated for Shearon Harris. The results of this evaluation showed no adverse effects on plant safety.
Safety valve blowdowns in excess of that assumed in the Shearon Harris FSAR will have the following effects on the events in which safety valve actuation occurs:
- l. Increased pressurizer water level during and following the valve blowdown,
- 2. Lower pressurizer pressure during and following valve blowdown,
- 3. Increased inventory through the valve.
The impact of the increased safety valve blowdowns with respect to the above effects was evaluated for the Shearon Harris FSAR events in which the safety valve actuation occurs (i.e., Loss of External Electrical Load and Single Reactor Coolant Pump Locked Rotor, and Feedwater System Pipe Break).
For the loss of External Electrical Load event, results from sensitivity analyses performed for a 4-loop plant were used for the evaluation. These analyses investigated the effects of different blowdown rates on the event. Similar results are expected for a 3-loop plant. The results of these analyses showed only marginal increases in pressurizer water volume and the maximum pressurizer water levels were well below the level at which liquid relief would occur. The Shearon Harris FSAR IBMD-PROC01-OS4 Question/Answer
QUESTION 6: Increased Blowdown RESPONSE: (continued) analysis results show that a small increase in pressurizer water volume, due to increased safety valve blowdown, would not result in liquid relief. The sensitivity analyses also showed that peak RCS pressures were unaffected by the increased blowdowns. The increased blowdowns did result in lower pressurizer pressure and increased RCS inventory loss, however, these had no adverse impact on the event and adequate decay heat removal was maintained.
For Single Reactor Coolant Pump Locked Rotor event, increased safety valve blowdowns have little impact. As analyzed and presented in the Shearon Harris FSAR, the opening and closing of the safety valve occurs over a short time period (less than 4 seconds). As a result, there is little change in either pressurizer level or RCS inventory. Increased safety valve blowdowns would have no impact on peak pressure, peak clad temperature, or minimum DNBR as these occur prior to the closing of the safety valve.
For the Feedwater System Pipe Break transient, the current FSAR analysis results show that the pressurizer safety valves open and remain open for the duration of the transient. Excess valve blowdown due to delayed closure of the safety valve will not affect the transient.
IBMD-PROC01-OS4 Question/Answer
Thermal expansion of the pressurizer and inlet piping to the valves will induce loading on the inlet flanges of the safety valves and PORVs at the time when the valves are required to lift. Additional loads due to fluid forces will be imposed Evaluate the effects that these when valve discharge begins.
loadings may have on valve operability. Give the maximum bending moment values predicted for the discharge flanges of the safety valves and PORVs resulting from the combined effects of earthquake, thermal expansion and pipe discharge forces.
Make a comparison of the predicted in-plant safety valve and PORV bending moments to the bending moments sustained by the test valves to demonstrate that the operability is not affected.
RESPONSE
The inlet loads induced on the safety and relief valves tested by EPRI exceed the loads predicted for the Shearon Harris Safety and Relief Valves. The maximum moment tested for the 6M6 valve was during test 908 and was 298.75 in-K. The largest moment predicted for the safety valve inlet at Shearon Harris is 248.80 in-K. This demonstrates functionability for the Shearon Harris safety valves.
Likewise, a bending moment of 43.0 in-K was induced in the inlet of the Copes<<Vulcan PORV test valve per EPRI 64-CV-174-2S. The largest moment predicted for the PORV inlet in the Shearon Harris valves is 30.97 in-K. This demonstrates functionability of the Shearon Harris relief valves.
Predictions have also been made for values of maximum bending moments on the discharge flanges. For the safety valves, the largest is 230.04 in-K. For the PORV's, the largest is 30.63 in-K.
IBMD-PROC01-OSR Question/Answer
The submittal did not provide an evaluation of the electric control circuits associated with the PORVs. NUREG-0737, Item II.D.1 requires the qualification of the control circuits of the plant specific PORVs for design-basis transients and accidents. However, such qualification does not have to submitted for this review, if it has already been included in submittal to fulfill the requirements of 10 CFR 50.49. Verify whether the in-plant PORV control circuits have been included in the 10 CFR 50.49 review or provide the necessary evaluation to demonstrate that the requirement of NUREG-0737, Item II.D.1 concerning control circuitry has been met, if the PORV circuitry was not reviewed under 10 CFR 50.49.
RESPONSE
The SHNPP Pressurizer PORVs actuation circuitries are not considered safety related because they are not assumed to open to mitigate the consequences of an accident. The PORVs are a control grade system and as such are not included in the 10 CFR 50.49 Qualification Program. The pressurizer PORVs will perform their intended function of remaining shut during an accident due to their design of failing shut upon loss of power. The potential for spurious actuation has been analyzed.
In addition, Control System failures have been analyzed in response to IE Notice 79-22. This analysis has been reviewed and approved by the NRC as documented in SER Section 7.7.2.2.
In addition, it should be noted that the PORVs have been qualified under the pump and valve operability program (PVORT),
the actuation pressure transmitters are environmentally qualified, the cable is qualified (thought not run as IE) and the PIC cabinets are essentially the same hardware as the Class 1E cabinets.
IBMD-PROC01-OSR t
Ques ion/Answer
The submittal transmitted a copy of the stress report entitled "Analysis of Pressurizer Power Operated Relief Valve and Safety Valve Discharge Piping" which presented the thermal hydraulic and stress analysis of the portion of the piping downstream of the PORVs and safety valves. The Class 1 piping upstream of the PORVs and safety valves was not addressed, Provide an evaluation of the thermohydraulic analysis of the Class 1 portion of the PORV and safety valve piping including but not limited to the following items:
a~ Provide detailed information on the program used so that the methodology for generating fluid discharge forces can be evaluated. Identify parameters such as timestep, valve flow area, pressure ramp rate, choked flow junction, and node spacing and discuss the rationale for their selection. Provide detailed information on how the program or methodology was verified for this application.
- b. Identify the program or methodology for calculating the fluid forces for the structural analysis. Discuss the accuracy of the results and the procedures used. to qualify the program or methodology. Provide a comparison of the calculated results using this program for an EPRI test condition with the EPRI test results for the same test.
C ~ Identify the initial conditions for the safety and relief valve thermal-hydraulic analyses. Describe the method used for treating valve resistance in the analyses and report flow rates corresponding to the resistances used.
Because the ASME Code requires derating of the safety valves to 90X of actual flow capacity, the safety valve analysis should be based on a flow rating equal to ill% of the flow rate stamped on the valve, unless another flow rate can be justified. Provide further information explaining how derating of the safety valves was handled and describing methods used to establish flow rates for the safety valves and PORVs in the thermal hydraulic analyses.
RESPONSE
The thermal hydraulic analysis presented in the report entitled "Analysis of Pressurizer Power Operated Relief Valve and Safety Valve Discharge Piping" (Reference 2) addresses downstream piping as well as the Class 1 piping upstream of the PORV and safety valve. Figures 3.1.1 and 3.1.2 of the report show the RELAP5 hydraulic model and the piping segments designation for forces, respectively. Therefore, the response to Question 9 is combined with the response to Question 10 and is presented below.
IB1$ -PROCO1-OSR Question/Answer
UESTION 10 Discharge Piping Thermal Hydraulic Analysis (Downstream of the PORV and Safety Valves)
The following questions are based on the review of the report, "Analysis of Pressurizer Power Operated Relief Valve and Safety Valve Discharge Piping" which was presented by the licensee as a part of the submittal.
a ~ The submittal indicated that thermal hydraulic analyses were performed for simultaneous actuations of the two PORVs in one analysis and the simultaneous actuation of three safety valves in another. It did not, however, verify that these analyses were performed using fluid conditions that produce maximum loading on the safety valve/PORV piping system. Provide evidence that the analyses were performed for transient conditions that produce maximum expected loading on the piping system.
Identify the fluid conditions assumed including pressure, temperature, pressurization rate, and fluid range type (steam, water, etc.).
- b. The submittal states that the thermal hydraulic analysis was performed using RELAP5/MOD1 and that forcing functions were calculated from RELAP5 output with the Code CALPLOTFIII. Provide verification that the latter code has produced accurate force histories for similar problems.
C ~ Identify input parameters used in the thermal hydraulic analysis such as timestep, valve flow area, pressure ramp rate, choked flow junction and node spacing and discuss the rational for their selection.
- d. Provide information explaining how the safety valve derating was handled in the analysis (refer to Question 9, Item c).
RESPONSE
In the course of testing, it was determined that the safety valve loop seal temperature profile may not be as hot as needed to cause the majority of the loop seal water to flash to steam as the slug traverses the discharge piping. This was the basis of the time-history forcing function input. To enhance the system, a slug trap arrangement was designed and implemented to divert the slug just downstream of each safety valve and, thus, preclude the slug from traversing all the discharge piping in the event the safety valves would open. To account for this, a series of statically equivalent forces were applied to the model simultaneously at each safety valve outlet using an appropriate dynamic load factor. These forces were considered along with the time-history forces in the system evaluat'ion.
IBMD-PROCOl-OSR Question/Answer
UESTION 10 RESPONSE:(continued) a ~ The limiting transient condition for Shearon Harris Nuclear Power Plant are provided in EPRI report "Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valves in Westinghouse-Designed Plants" (Reference 4) and are shown in Table 2.1 of the submittal entitled "Analysis of Pressurizer Power Operated Relief Valve and Safety Valve Discharge Piping" (Reference 2).
The limiting event resulting in SRV steam discharge is the Locked Rotor Case. Analyses were performed with the pressurizer pressure at 2559.55 psig (2485 psig set point
+ 3X accumulation) with zero pressurization rate and steam at saturation temperature. Three cases with average loop seal temperatures of 194'F, 310'F, and 367'F were analyzed as well as the case of a drained loop seal.
The limiting event resulting in SRV liquid discharge is the Hain Feedline Break, with water temperature in the 620.1 to 623.4'F range. This case was not analyzed as it is bounded by the loop seal case since significant flashing of the liquid would occur subsequent to valve opening. EPRI tests 1017 and 1027 (Reference 3) demonstrated the lower loading for the liquid discharge.
The limiting event resulting in PORV steam discharge is the Locked Rotor case. The analysis was performed for a cold loop seal followed by saturated steam from a pressurizer at 2405.05 psig (2335 psig set point + 3%
accumulation) .
The limiting event resulting in PORV liquid discharge is actuation by the cold overpressure protection system. To ensure that the above is the bounding case, the loading on the piping system was developed for the high pressure LTOP case (pressurizer at 2335 psig + 3X and liquid at 280'F).
The loading was less than in the loop seal case.
The high pressure injection at power event is bounded by the above two cases for both first (steam) and subsequent opening (liquid) since the pressurizer is at a lower pressure.
- b. The post-processor CALPLOTFIII was performed to convert the transient flow conditions (calculated by RELAP5/MOD1) into transient forces on the piping system. The derivation of the governing equations are shown in Appendix A of EBASCO report (Reference 2). The validity of the program coding was verified by comparing hand calculation results against the values computed by the program. The program was further assessed against the GE 4-inch pipe blowdown test results. Favorable comparisons were obtained in comparing the computed results against the test data.
IBMD-PROC01-OSR Question/Answer
UESTION 10 RESPONSE:(continued)
CALPOTFIII was also verified by running CE test 1411 for SRV actuation on RELAP5/MOD1 using the input, from EPRX's RELAP5/MOD1 application (Reference 3). The calculated hydrodynamic conditions were converted by CALPLOTFIII to transient forces that duplicated the forces obtained by EPRI (Reference 3).
- c. The following are the input parameters used in the thermal hydraulic analysis:
- i. The node spacing varied from 0.75 ft near the valves to 2.0 ft. in the downcomer. However, on average the volume length was 1.0 ft.
-5 ii. The maximum time step used was 5x10 seconds for the SRV analysis and 10 seconds for the PORV analysis.
This selection satisfies the criteria that no front (whether pressure or fluid) may traverse the length of a control volume in one time step.
iii. The valve flow areas used were 0.02040 and 0.03425 sq. ft. for the SRV and the PORV, respectively. The SRV flow area was adjusted since the full area of 0.0253 sq. ft. results in flow much higher than the valve rated flow (see response to 10(d)). The PORV flow area used was in error and should have been 0.017 sq. ft., and thus produces lower loads than those used in the stress analysis.
iv. The choking option was used only at the valve junctions as recommended by EPRI (Reference 3).
- d. The PORV was modeled by the "MRTVLV" option in RELAP5/MOD1, with linear opening of the valve area. A table of Cv vs. stem position specifies the resistance of the valve.
The SRV is also modeled by the "MTRVLV" option, but because of the short opening of the valve and following EPRI's application of RELAP5/MOD1 (Reference 3) the resistance is represented by the abrupt area change option of the code. The CROSBY 6M6 valve has a flow area of 0.0253 sq. ft. and rated flow capacity of 420,000 ibm/hr.
However, as was shown in EPRI's application of RELAP5/MOD1 it is necessary to reducesteady-state the valve area to 0.0204 sq. ft.
flow rate. The SRV to achieve the measured flows rate calculated by the hydraulic analysis, using the adjusted area is 499,320 ibm/hr which is 119X of the flow
~
rating of valve and thus satisfies the ASME code requirement of 90X derating.
IBMD-PROC01-OS4 Question/Answer
UESTION 11 As stated in Question 9 above, the submittal did not present the stress evaluation of the Class 1 (inlet) portion of the PORV and safety valve piping and supports. Provide a desc'ription of the Class 1 piping analysis and stress evaluation including, but not limited to the following items:
a~ A description of the method and procedure used to perform the stress analysis. Identify the computer programs used and provide verification to demonstrate that the program produces accurate results for this type of analysis.
- b. A description of methods used to model supports, the pressurizer and relief tank connections, and the safety valve bonnet assemblies and PORV actuator.
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. Identify the mathematical methods used to perform the load combinations, and the governing codes and standards used to determine piping and support adequacy.
RESPONSE: (a)
The analytical methods used to obtain a piping deflection solution consist of the transfer matrix method and stiffness matrix formulation for the static structural analysis. The response spectrum method is used for the seismic dynamic analysis.
The complexity of the piping system requires the use of a computer to obtain the displacements, forces, and stresses in the piping and support members. To obtain these results, accurate and adequate mathematical representations (analytical models) of the systems are required. The modeling considerations depend upon the degree of accuracy desired and the manner in which the results will subsequently be interpreted and evaluated. All static and dynamic analyses are performed using the WESTDYN computer program. This program, described in WCAP-8252, was reviewed and approved by the U.S.
NRC (NRC letter, April 7, 1981 from R. L. Tedesco to T. M.
Anderson).
The integrated piping/supports system mode is the basic system model used to compute loadings on components, component and piping support, and piping. The system model includes the stiffness and mass characteristics of the piping, attached equipment, and the stiffness of supports, which affects the system response. The deflection solution of the entire system is obtained for the various loading cases from which the internal member forces and piping stresses are calculated.
IBMD-PROCOl-OSR Question/Answer
UESTION 11 RESPONSE: (a) (continued)
Static Analysis The piping system models, constructed for the WESTDYN computer program, are represented by an ordered set of data, which numerically describes the physical system.
The spatial geometric description of the piping model is based upon the isometric piping drawings and from equipment drawings referenced in the design specification. Node point coordinates and incremental lengths of the members are determined from these drawings. Node point coordinates are put'n network cards. Incremental member lengths are put on element cards.
The geometrical properties along with the modulus of elasticity, E, the coefficient of thermal expansion, ~, the average temperature change from the ambient temperature, AT, and the weight per unit length, w, are specified for each element. The supports are represented by stiffness matrices which define restraint characteristics of the supports.
The static solutions for deadweight and thermal loading conditions are obtained by using the WESTDYN computer program.
The WESTDYN computer program is based on the use of transfer matrices which relate a twelve-element vector (B) consisting of deflections (thre'e displacements and three rotations) and loads (three forces and three moments) at one location to a similar vector at another location. The fundamental transfer matrix for an element is determined from its geometric and elastic properties. If thermal effects and boundary forces are included, a modified transfer relationship is defined as follows:
11 12 F
21 22 o~)
or T1B1 o
+R ~B where the T matrix is the fundamental transfer matrix as describe above, and the R vector includes thermal effects and body forces. This B vector for the element is a function of geometry, temperature, coefficient of thermal expansion, weight-per unit length, lumped'asses, and externally applied loads.
The overall transfer relationship for a series of elements (a section) can be written as follows:
IBMD-PROC01-OSR Question/Answer
UESTION 11 RESPONSE: (a) (continued)
~ T1B + R1 B1 1 o 2 21 2 21o 21 2 2
= T B2 + R3 + T T2R1 + T +
B3 3220 T T T2B R R or n n B=mT+B+E n 1 1 0 7f 7
T r R r-1 + R n
A network model is made up of a number of sections, each having an overall transfer relationship formed from its group of elements. The linear elastic properties of a section are used to define the characteristic stiffness matrix for the section.
Using the transfer relationship for a section, the loads required to suppress all deflections at the ends of the section arising from the thermal and boundary forces for the section are obtained. These loads are incorporated in the overall load vector.
After all the sections have been defined in this manner, the overall stiffness matrix, K, and associated load vector needed to suppress the deflection of all the network points is determined. By inverting the stiffness matrix, the flexibility is determined. The flexibility matrix is multiplied by 'atrix the negative of the load vector to determine the network point
'eflections due to the thermal and boundary force effects.
Using the general transfer relationship, the deflections and internal forces and then determined at all node points in the system. The support loads, F, are also computed by multiplying the'tiffness matrix, K, by the displacement vector, 5, at the support point.
Dynamic Analysis The models used in the static analyses are modified for use in the dynamic analyses by including the mass characteristics of the piping and equipment.
Seismic Analysis The lumping of the distributed mass of the piping systems is accomplished by loading the total mass at points in the system which will appropriately represent the response of the distributed system. Effects of the equipment motion, that is, the pressurizer, on the piping system are obtained by modeling the mass and the stiffness characteristics of the equipment in the overall system model.
lBMD-PROC01-OSR Question/Answer
UESTION 11 RESPONSE: (a) (continued)
The supports are again represented by stiffness matrices in the system model for the dynamic analysis. Mechanical shock suppressors which resist rapid motions are now considered in the analysis. This solution for the seismic disturbance employs the response spectra method. This method employs the lumped mass technique, linear elastic properties, and the principle of modal superposition.
From the mathematical description of the system, an overall stiffness matrix (K) is developed from the individual element stiffness matrices using the transfer matrix (k ) associated with mass degrees-of-freedom only. From the mass matrix and the reduced stiffness matrix, the natural frequencies and the normal modes are determined. The modal participation factor matrix is computed and combined with the appropriate response spectra value to give the modal amplitude for each mode. Since the modal amplitude is shock direction dependent, the total modal amplitude is obtained conservatively by the absolute sum of the contributions for each direction of shock. The modal amplitudes are then converted to displacements in the global coordinate system and applied to the corresponding mass point.
From these data the forces, moments, deflections, rotation, support reactions, and piping stresses are calculated for all significant modes.
The seismic response from each earthquake component is computed by combining the contributions of the significant modes.
Thermal Transients Operation of a nuclear power plant causes temperature and/or pressure fluctuations in the fluid of the piping system. The transients for this system are defined in "Westinghouse Systems Standard Design Criteria 1.3" and referenced in the design specification. These were used to define the various operating modes used in the thermal expansion analyses.
Valve Thrust Analysis The safety and relief lines were modeled statically and dynamically. The mathematical model that is used in the seismic analysis was modified for the valve thrust analysis to represent the safety and relief valve discharge. The time-history hydraulic forces supplied by Ebasco were applied to the piping system lump mass points. The dynamic solution for the valve thrust was obtained by using a modified-predictor-integration technique and normal mode theory.
IBMD-PROC01-OSR Question/Answer
UESTION 11 RESPONSE: (a) (continued)
The time-history solution was found using program WESTDYN by first determining the natural frequencies and normal modes of the pressurizer safety and relief line model utilizing the applied hydraulic loads. The resulting time-history displacements were then used to calculate internal forces and deflections at each end of the piping elements. For this calculation, the displacements were tieated as imposed deflections of the pressurizer safety and relief line masses.
The time-history internal forces and displacements are then used to determine the maximum forces, moments, and displacements that exist at each end of the piping elements and the maximum loads for piping supports. The results from program WESTDYN are saved on TAPE14 for future use in piping stress analysis and support load evaluation.
The equivalent static safety valve thrust analysis was also performed as discussed in Question 9. These loadings were enveloped with those determined by time-history analysis. This total loading was used as the safety valve thrust loading for the system.
RESPONSE: (b)
The supports are represented by stiffness matricies which define restraint characteristics of the supports.
The pressurizer nozzle connections were modeled as equivalent pipe sections which developed the proper stiffness effects.
Mass effects were also determined and included in the model.
The relief tank nozzle connections have been modeled as actual pipe cross section. The reinforcement pad was modeled applying the appropriate shell stiffness in the bending moment directions. The shear and axial force directions as well as the torsional moment directions were modeled as rigid restraints.
The safety valve bonnet assemblies and PORV actuators were modeled by applying a lump mass equivalent to the valve body and actuator at the valve center of gravity (CG) location.
Representative pipe properties were utilized.
IBMD-PROC01-OSR Question/Answer
QUESTION 11 RESPONSE (c) (continued)
In order to evaluate the pressurizer safety and relief valve piping, appropriate load combinations and acceptance criteria were developed. The load combinations and acceptance criteria include those recommended by the piping subcommittee of the. PWR PSARV test program and are outlined in Tables 1 and 2 with a definition of load abbreviations provided in Table 3.
IBMD-PROC01-OSR Question/Answer
Table 1 Load Combinations and Acceptance Criteria for Pressurizer Safet and Relief Valve Pi in U stream of Valves Piping Plant/System Allowable Stress
.0 eratin Condition Load Combination Xntensit Normal 1.5 S m
Upset N + OBE + SOTU 1.8 S /1.5 S m y Emergency N + SOT 2.25 S m
/1.8' S Faulted N + SSE + SOTF 3.0 S m
NOTES: (1) See Table 3 for SOT definitions and other load abbreviations.
(2) Use SRSS for combining dynamic load responses.
(3) Use absolute summation of static and dynamic responses.
IBMD-PROC01-OSR Question/Answer
Table 2 Load Combinations and Acceptance Criteria for Pressurizer Safet and Relief Valve Pi in - Downstream of Valves Piping Plant/System Allowable Stress 0 eratin Condition Load Combination Intensit Normal 1.0 Sh Upset N + OBE + SOT 1.8 Sh Emergency N + SOT 1.8 Sh Faulted N + SSE + SOT 2.4 Sh NOTES: (1) See Table 3 for SOT definitions and other load abbreviations.
(2) Use SRSS for combining dynamic load responses.
(3) Use absolute summation of static and dynamic responses.
IBMD-PROC01-OSR Question/Answer
Table 3 Definitions of Load Abbreviations Sustained loads during normal plant operation SOT System operating transient SOT Relief valve discharge transient SOTE Safety valve discharge transient SOT Max (SOTU, SOTE); or transition flow OBE Operating basis earthquake SSE Safe shutdown earthquake Basic material allowable stress at maximum (hot) temperature S Allowable design stress intensity m
S Yield strength value IBMD-PROC01-OSR Question/Answer
The method used to combine the primary loads to evaluate the adequacy of the piping system is described below.
Primary Stress Evaluation In order to perform a primary stress evaluation in accordance with the rules of the Code, definitions of stress combinations are required for the normal, upset, emergency and faulted plant conditions. Tables 1 and 2 illustrate the allowable stress intensities for the appropriate combination. Table 3 defines all pertinent terms.
Design Conditions The piping minimum wall thickness, t , is calculated in accordance with the Code. The actual pipe minimum'wall Vhickness meets the Code requirement.
The combined stresses due to primary loadings of pressure, weight, and design mechanical loads calculated using applicable stress intensity factors must not exceed the allowable limit. The resultant moment, M., due to loads caused by weight and design mechanical loads is calculated using the following equation:
M + M + M + M DML yDML 1/2
+ M + M wt DML where M
xwt '> M 'I M = deadweight moment components wt wt DML DML DML design mechanical load moment components Upset Conditions The combined stresses due to the primary loadings of pressure, weight, OBE seismic, and relief valve thrust loadings calculated using the applicable stress intensity factors must not exceed the allowables. The resultant moments, M , due to loads caused by these loadings are calculated as shown below.
IBMD-PROC01-OSR Question/Answer
For seismic and relief valve thrust loading:
1/2 2 2
M + M + M + M OBE SOTU yOBE ySO U
1/2 2
+ M + M +
wt OBE SOT Where:
M , M , M deadweight moment comp'onents wt wt wt M , M , M internal OBE moment components OBE 5BE OBE M , M , M relief line operation moment components SOT ySOT SOT U U U Emergency Conditions The combined stresses due to primary loadings of pressure, weight and safety valve thrust, using applicable stress intensification factors, must not exceed the allowable limits. The magnitude of the resultant moment, M. is i calculated from the moment components as shown below:
1/2 2
M X
+ M x + M + M + M + M SO E
wt SOTE wt Where:
M M ', M ~ deadweight moment components X y wt wt wt M,
x '
SOT M, SOTE
' M SOT safety line operation moment components IBMD-PROC01-OSR Question/Answer
Faulted Conditions The combined stresses due to primary loadings of pressures, weight, SSE and SOT , using applicable stress intensification factors must not exceed the allowable limits. For the resultant moment loading, M , the SSE and SOTF moments are combined using the square-root-of-the-sum-of-the-squares (SRSS) addition and added absolutely with deadweight for each moment component (M , M , M ).
The magnitude of the resultant moment, M , is calculated'romm tRe three moment components, as shown below:
2 2 1/2 M
i M SOT F
+ M SSE
+ M xwt 2 2 1/2 M ~ M M ySOT ySSE ywt F
2 2 1/2 1/2 M + M + M z
SOT SSE wt F
where M M M deadweight moment components xwt ywt z wt M, M, SSE ySSE M
SSE inertial SSE moment components M
M,'
y M maximum components of SOTU or SOTE moment SOTE SOTE SOTE For the safety and relief piping, the faulted condition load combination of pressure, weight, and valve thrust is considered as given in Tables 1 and 2 and defined in Table 3. The pipe break loads (Main Steam, Feedwater, or LOCA) can be ignored for the PSARV system. These loads have very little impact on the pressurizer safety and relief system when compared to the loading conditions discussed here.
IBMD-PROC01-,0SR Question/Answer
Secondary Stress Evaluation The combined stresses due to the secondary loadings of thermal, pressure, and deadweight using applicable stress intensification factors must not exceed the allowable limit. Por the resultant moment loading, Mi, thermal moments are combined as shown below:
2 -M M -M + M =
-M + M 1/2 x x Z z 2 MAX MIN MAX MIN MAX MIN M M M maximum thermal moment considering all MAX yMAX MAX thermal cases including normal operation M M M Z
minimum thermal moment considering all MIN y>>IN MIN thermal cases including normal operation This M , is then substituted into the appropriate equations of the applicable code.
H The governing codes and standards used to determine piping adequacy are:
- 1. Class 1 Piping ASME Section III - Subsection NB, 1971 Edition and all edition and addenda through the Summer 1979 Addendum.
- 2. NNS Piping ANSI B31.1 "Power Piping" 1973 Edition through Summer 1975 Addenda.
IBMD-PROCOl-OSR Question/Answer
Support Load Combinations The support load combinations developed for use in the support design were based on criteria supplied by Ebasco via letter EB-C-19464 9/16/85.
For NORMAL Loading Larger of: a. MAX(THERM NORM) + DWT
- b. MIN(THERM NORM) + DWT
- c. DWT For UPSET Loading Larger of: a. MAX(THERM) + DWT
- b. MIN(THERM) + DWT
For EMERGENCY LOADING Larger of: a. MAX(THERM) + DWT
- b. MIN(THERM) + DWT
For FAULTED Loading Larger of: a. MAX(THERM) + DWT
- b. MIN(THERM) + DWT
- c. DWT plus: SRSS of (SSE and SVT)
Where:
Deadweight Loading MAX(THERM NORM) Highest Positive Thermal Loading for all normal operation thermal cases.
MIN (THERM NORM) Highest Negative Thermal Loading for all normal operation thermal cases.
MAX(THERM) Highest Positive Thermal Loading for all thermal loading conditions MIN(THERM) Highest Negative Thermal Loading for all Thermal Loading Coditions OBE Operation Basis Earthquake RVT Relief Valve Thrust SSE Safe Shutdown Earthquake (also known as DBE)
SVT Safety Valve Thrust IBMD-PROC01-OSR Question/Answer
REFERENCES
- 1. "Report on the Operability of Pressurizer Safety Relief Valves, Power Operated Valves and Electrical Motor Operated Block Valves for Carolina Power and Light Company: Shearon Harris Nuclear Power Plant", April 1984.
- 2. "Analysis of Pressurizer Power Operated Relief Valve and Safety Valve Discharge Piping for Shearon Harris Nuclear Power Plant",
April 1984.
- 3. Application of RELAP5/MODl for Calculation of Safety and Relief Valves Discharge Piping Hydrodynamic Loads" EPRI NP-2479, December 1982.
- 4. "Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valves in Westinghouse Designed Plants" EPRI NP-2296, December 1982.
- 5. "Safety and Relief Valve Test Report" EPRI NP-2628-SR, December 1982 IBMD-PROC01-OSR Question/Answer