ML20199L912
| ML20199L912 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway, 05000000 |
| Issue date: | 06/30/1986 |
| From: | Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM SLNRC-86-07, SLNRC-86-7, NUDOCS 8607100056 | |
| Download: ML20199L912 (10) | |
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SNUPPS Standardized Nuclear Unit Power Plant System 5 Choke Cherry Road Nicholas A. Petrick Rockville, Merytand 20850 Executive Director June 30, 1986 SLNRC 86-07 FILE: 0278 SUBJ:
NUREG-0737, Item II.D.1 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation
~U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Docket Nos.: STN 50-482 and,TN 50-483 S
References:
- 1. NRC letter (P. O'Connor) to Union Electric Company (D. Schnell) dated April 22, 1986: Request for Additional Information NUREG-0737, Item II.D.1 Performance Testing of Relief and Safety Valves
- 2. NRC letter (P. O'Connor) to Kansas Gas & Electric Company (G. Koester) dated April 22, 1986: Request for Additional Information NUREG-0737, Item II.D.1 Performance Testing of Relief and Safety Valves
Dear Mr. Denton:
The referenced letters requested that additional information be provided in support of the NRC review of NUREG-0737, Item II.D.1 for the SNUPPS plants--
Callaway Plant and Wolf Creek Generating Station.
Enclosed is a partial response to the NRC staff questions. Of the 15 questions in the requests for additional information, the enclosure provides responses to 9 questions. The' responses to the remaining questions require additional time to prepare.
It is anticipated that these responses will be provided by September 26, 1986. You will be notified if there are any problems with meeting this date.
Ver truly yours, t
4.%c cholas A. Petrick MHF/dck/7a3
Enclosure:
Partial Response to' Request for Additional Information cc: D. F. Schnell UE G. L. Koester KGE J. M. Evans KCPL B. Little USNRC/ CAL J. E. Cummins USNRC/WC W. L. Forney USNRC/RIII E. H. Johnson USNRC/RIV
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o Responses to NRC Questions re NUREG-0737, Item II.D.1 Performance Testing of Relief and Safety Valves Question 1:
In valve operability discussions on cold overpressurization transients, the submittal only identifies conditions for water discharge transients.
According to the Westinghouse valve inlet fluid conditions report, however, the PORVs are expected to operate over a range of steam, steam-water, and water conditions because of the potential presence of a steam bubble in the pressurizer and water solid operations. To assure that the PORVs operate for all cold overpressure events, discuss the range of fluid conditions expected for the expected types of fluid discharge and identify the test data that demonstrate operability for these cases.
Since no low pressure steam tests were performed for the PORVs, confirm that the high pressure steam tests demonstrate operability for the low pressure steam case for both opening and closing of the PORVs.
Response
As discussed in EPRI NP-2296, the PORVs may~ experience a range of steam, steam-water and water conditions during mitigation of cold overpressure events.
The SNUPPS plants use Garrett PORVs. Section 3.11 of EPRI NP-2460-SR dis-cusses the Garrett PORV and notes that the testing was designed.to demon-strate operability under steam, transition and water conditions. The Garrett PORV is designed to use the differential pressure across the valve, from inlet to outlet, for opening the valve against the spring force which holds the valve plug closed.
Differential pressure is also used to close the valve by assisting the plug spring force. Appropriate routing of the pressures within the valve is controlled by a solenoid-operated pilot valve.
Orifices within the valve body control the opening and closing rate of the valve plug.
Although no low pressure steam tests were performed, the proper functioning t
of the PORVs under reduced pressure water tests, for pressures ir the 495 psia to 683 psia range, provides assurance that the valves will operate under low pressure steam conditions and low pressure transition conditions as well as low pressure water conditions.
i Based on the above, the tests identified in EPRI NP-2460-SR Table 3-31 l
provide assurance of proper PORV function throughout the range of inlet conditions for the SNUPPS plants.
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Question 2:
Results for the EPRI tests on the Crosby safety valves indicate that the test blowdowns exceeded the design value of 5% for both "as installed" and " lowered" ring settings.
If the blowdowns expected for SNUPPS also exceed 5%, 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 sufficiently e
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decreased such that adequate cooling might not be achieved for decay heat removal. Discuss these consequences of higher blowdowns if increased blowdowns are expected.
Response
The response to this question is scheduled for 9/26/86.
Question 3:
The plant specific submittal stated that the loop seal is to be insulated to bring the valve inlet loop seal temperature to 300*F. EPRI Tests 1415 and 1419 tested the Crosby 6M6 valve with loop seal inlet temperatures of 290 F and 350 F, respectively. Test 1415 demonstrated the valve fluttered or chattered during loop seal discharge, opened on steam and stabilized, and closed smoothly. Test 1419 performed similarly with the exception that the test was terminated after observing valve chatter during closure.
Discuss the analysis performed to estimate the valve inlet temperature.
Confirm that the loop seal temperature presented in the submittal is repre-sentative of Test 1415 and that the phenomena observed for Test 1419 is not applicable to the SNUPPS plants. Will field measurement of the loop' seal temperature be performed to verify the analysis?
If so, provide field measurements when available.
Response
The loop seals for the SNUPPS plant safety valves have been insulated to maintain an elevated fluid temperature at the valve inlet. The amount of insulation which was installed was determined by analysis.. The analysis assumed that the loop seal / steam interface is maintained at a constant temperature of 652.7 F.
A heat balance was performed by calculating the quantity of heat which must be dissipated from the loop seal, via conduction and convection, to achieve a temperature of approximately 350 F at the valve inlet. This quantity of heat is set equal to the amount of heat which must be transferred through the piping wall and insulation.
From the heat balance, the required amount of insulation is determined.
It was conser-vatively estimated, as a result of this analysis, that the water temperature i
at the valve inlet would be greater than 300 F which was the temperature used in the analysis of safety valve piping.
P., in-plant measurement of loop seal temperatures has been performed.
The' loop seal temperature at the valve inlet is not expected to affect I
the stability of the safety valve on closure. WCAP-10105, section 3.4, I
discusses valve chatter upon closure and relates this phenomenon to pressure pulses in the inlet piping.
The frequency of inlet piping pressure pulses is a function of inlet piping length and steam sonic velocity. Therefore, l
the temperature of the loop seal water is not expected to be a factor on closure since the loop seal -fluid would have cleared and the inlet piping would be filled with steam. The high temperature loop seal does result in the known benefit of lower dynamic loads on the valve and downstream piping during loop seal clearing.
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Question 4:
The plant specific submittal identifies the Westinghouse Inlet Fluid Condi-
.tidns Report which provides the bounding inlet fluid conditions for a ref--
erence 4-loop plant. However, backpressure at the valve discharge was not provided.
Since backpressure affects valve performance, provide the expected SNUPPS backpressure.
Identify the EPRI tests applicable to the SNUPPS safety valves which are representative of the expected inlet conditions.
Response
Backpressures measured in EPRI Tests of the 6M6 Crosby Safety Valve varied from 245 to 725 psia.
For SNUPPS Safety Valves, the maximum backpressure is expected to be 481 psia which is well within the range measured by EPRI.
However, EPRI stated in' Report NP-2770-LD Volume 6 that Crosby valve per-formance was relatively unaffected by back pressure. This is due to the bellows design of the Crosby valve. Therefore, the specific tests with
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backpressure applicable to the SNUPPS plants are tests 929, 931, 932, 1406, 1411, 1415 and 1419 from Table 3.5.1.b of NP-2628-SR.
Question 5:
The EPRI Inlet Fluid Justification Report suggested a method for demon-strating safety valve stability. This method compares the total inlet piping pressure drop for the in-plant safety valves and piping to the appli-cable EPRI test safety valve and piping combinations. The total inlet piping pressure drop is composed of a frictional and acoustic wave component evaluated under steam conditions. The SNUPPS plant submittal did not pro-vide pressure drop calculations or any other methods to demonstrate safety valve stability. Provide the necessary documentation and discussion demon-strating stability for the SNUPPS plant safety yalves at the expected inlet conditions, ring settings and inlet piping configuration.
Response
The following table compares the EPRI (6M6) test piping configuration from EPRI NP-2628-SR, Table 3.5.1.a with the SNUPPS safety valve inlet piping arrangement.
EPRI SNUPPS Length, in.
I.D.,in.
Length, in.
I.D.,i n.
Nozzle 17 6.813 NA
. Venturi 38 6.813 NA Pipe 13 6.813 NA Reducer 6
6.813/4.897 NA Loop Seal
-straight 48 4.897 56 5.187
-bends 2-180 9 in. radius 1-180,2-90 9 in, radius Inlet Flange 19 4.897 10 5.187,
The EPRI-calculated transient pressure drop, which is a function of fluid friction losses and acoustic wave losses, for the test configuration is 251 psi. The SNUPPS inlet piping is approximately 5 feet shorter in length than the EPRI test configuration piping and the calculated corresponding SNUPPS pressure drop is 242 psi.
Since the SNUPPS piping length is less than the
. test configuration piping length and the SNUPPS pressure drop is less than the test configuration pressure drop, the SNUPPS safety valve performance should be more stable than the test valves with " reference" ring settings, i.e. tests 929, 931, 932, 1406, 1411, 1415 and 1419. These tests address the range of inlet fluid conditions as discussed in EPRI NP-2460-SR, Sec-tion 4.7.1.
Question 6:
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 PORV flow is also predicted for the cold over pressurization event. These same flow conditions will also exist for the PORV block valve.
The EPRI/ Marshall 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 PORV block valve could be expected to operate with liquid flows, discuss PORV block valve operability with expected liquid flow conditions and provide specific test data.
Response
The SNUPPS PORV block valves are Class 1E motor-operated valves and are environmentally and seismically qualified to perform their function for all design basis accidents and transients. The valves are the Westinghouse Model 3GM88 gate valve which was successfully tested in the EPRI Marshall test program after adjustments were made to the operator torque switch.
The EPRI-Marshall testing used saturated steam as the test
- fluid.
In follow-up testing after the EPRI-Marshall tests, Westinghouse performed operability testing on several gate valves including the Model 3GM88 with an SB-00-15 Limitorque operator.
The Westinghouse testing employed subcooled water as the test fluid.
Section 4.2.3 of the Westinghouse Gate Valve Closure Testing Program Summary Report discusses successful testing of the 3GM88 at differential pressures ranging from 80') to 2600 psi and flow rates from 60 to 600 gpm. The Westingh,ouse testing demonstrated that the torque required to operate the block valves is almost entirely dependent on the valve differential pressure.
Expected differential pressure's on the block valve at closure during plant transients involving water relief would be less than the EPRI-Marshall steam testing differential pressures (2450 psi) since the PORV actuation occurs at a pressure of 2335 psig.
In addition, water is a better lubricant than steam; thus, the valve inter-nal friction, which must be overcome by the valve motor operator, would be lower under liquid flow conditions.
Therefore, the operability demonstrated.
during the EPRI-Marshall testing may be compared against liquid flow condi-l tions with other factors (e.g., differential pressure) being the same.
Furthermore, as discussed in the July 24, 1981 letter from R.C. Youngblood, on behalf of PWR owners, to H.R. Denton, additio.nal block valve testing (for
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liquid water flow) is not necessary because of the low probability of a PORV failing to function properly, the operability of block valves is not a safety issue because current plant procedures can deal with this event and the EPRI-Marshall tests provide sufficient operability data.
Based on the above, sufficient testing has been performed to confirm block valve operability with l~iquid flow conditions.
Question 7:
Bending moments are induced on the safety valves and PORVs during the time they are required to operate because of discharge loads and thermal expan-sion of 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.
Response
The response to this question is scheduled for 9/26/86.
Question 8:
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 EPRI 6M6 test safety valve experienced some chatter and flutter while discharing liquid at cer-tain ring settings. Testing was terminated after observing chattering to minimize valve damage.
Inspection revealed some valve damage which was presumably caused by the valve chatter'and flutter. Liquid discharge for the SNUPPS plants may conceivably occur for longer periods of time than the EPRI testing. Thus, longer period of valve chattering may cause severe valve damage. Discuss the implications this may have on operability and reliability of the SNUPPS plant safety valves.
Identify any actions that will be taken to inspect for valve damage following safety valve lift events.
Response
The response to this question is scheduled for 9/26/86.
Question 9:
NUREG-0737 ITEM II.D.1 requ' ired that the plant-specific PORV control cir-cuitry be qualified for design-basis transients and accidents.
Please provide information which demonstrates that this requirement has been fulfilled..
e e
Response
The pressurizer PORVs and control circuitry are Class 1E equipment and
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are environmentally and seismically qualified for design basis transients and accidents in accordance with sections 3.11 and 3.10 of the SNUPPS plant FSARs.
Question 10:
The SNUPPS 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 details discussing the applicable EPRI tests which demon-strates the operability of the plant safety valves. The submi.ttal did not provide the present SNUPPS plant safety valve ring settings.
If the plant current ring settings were not used in the EPRI tests, the results may not be directly applicable to.the SNUPPS safety valves.
Identi fy the SNUPPS 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 spec-ific ring settings were extrapolated or calculated from the EPRI test data.
Identify t.hese values so determined and evaluate the effects of these' values on the behavior. of the safety valves.
Identify the applicable EPRI tests representative of these ring settings.
Response
The safety valve ring settings for the SNUPPS plants are as follows:
Plant Serial No.
N.R.
G.R.
Wolf Creek N60446-00-0001
-18
-260 0002
-18
-265 0003
-18
-230 Callaway N60FS-00-0004
-18
-230 0005
' 18
-230 0006
-18
-225 When the above ' guide ring settings are' adjusted to account for the " level" position, which was used by EPRI as the baseline for ring settings, the guide ring settings fall in the range of -60 to -120 notches. These values envelope the EPRI test ring settings for the Crosby valve tests with " refer-ence" ring positions as identified in EPRI NP-2460-SR, Table 4-7.
Therefore, SNUPPS valve performance should be similar to the EPRI tests with " reference" ring positions.
(It is noted that one of the test valves. test 1419 -
l exhibited chatter on closure. This stability concern is addressed under question number 5 above.)
Question 11:
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The Westinghouse 3GM88 EPRI test valve with a Limitorque SB-00-15 actuator successfully opened and closed on. command only after the torque switch was I
set to the maximum of 3.75 plus. The plant specific submittal stated that Westinghouse modified the 3GM88 block valve at the SNUPPS plants to provide sufficient closing thrust. The details of the block valve modification were not provided.
Provide a discussion of the modification to the 3GM88 block i
valve and the torque switch setting. _
Response
The block values at the SNUPPS plants were modified by Field Change Notices issued by Westinghouse. The modifications included adjusting the torque switch settings to optimal values for both opening and closing thrust. The Field Change Notices are available for review at the SNUPPS plants. At Callaway Plant, the torque switch settings have been verified and adjusted to design specification requirements through valve signature analysis type testing. The block valves at the SNUPPS plants use SB-00-15 actuators.
Question 12:
The Westinghouse WCAP-10105 report states that for the feedwater line break and extended high pressure injection events, liquid discharge only occurs after the pressurizer is liquid filled such that water reaches the safety valve inlets. The length of time to fill the pressurizer is plant dependent and varies from 20 min to 6 h.
Estimate the time to fill the SNUPPS plants' pressurizer based on the worst case of the two events. Since the safety valves were originally specified and designed for steam service only, discuss the effects of liquid discharge on valve operability. Provide a discussion addressing the pressurizer fill time and if there is sufficient time for the operators to take corrective action to prevent liquid discharge through the safety valves for these events.
Response
The response to this question is contingent upon the response to question 8 above which will be provided at a later date.
Question 13:
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) Evidence that the analysis was performed on the fluid transient cases producing the maximum loading on the safety /PORV piping system.- The cases should bound all steam, steam to water, and water flow transient conditions for the s'afety and PORY valves.
(b)
Identiification of important parameters used in the thermal hydraulic analysis and rationale for their selection. These include peak pres.
sure and pressurization rate, valve opening time, fluid conditior et valve opening, time step and valve flow area.
(c) A sketch of the thermal hydraulic model showing the size and number of fluid control volumes.
Response
The response to this question is scheduled for 9/26/86.
Question 14:
The submittal states that a structural analysis of the safety PORV valve piping system has been conducted. To allow for a more complete evaluation of the methods used and the results obtained from the structural analysis, please provide, reports containing at least the following information:
(a) An evaluation of the results of the piping support analyses including identification of overstressed locations and a description of modifi-cations, if any.
(b)
Identification of important parameters used in the structural analysis and the rationale for their selection. These include node spacing, time step, damping and cut off frequency.
(c) A sketch of the structural model showing lumped mass locations, pipe sizes, and application points of fluid forces.
Response
The response to this question is scheduled for 9/26/86.
Question 15:
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 prior 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.
Provide the peak pressures expected at the SNUPPS plants and a comparison to the pressure allowed in WCAP 10105. The pressure oscillations could potentially excite high frequency vibration modes in the piping, creating bending moments in the inlet piping that should be combined with moments from other appropriate mechanical loads.
Provide one of the following: (1) a comparison of the expected peak 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.
R'esponse:
The acoustic pressures in the safety valve inlet piping were measured d,uting EPRI testing of Crosby safety valves and were analyzed in WCAP-10105. ' The EPRI testing did not result in piping damage, and the analysis showed that expected acoustic pressures are below the maximum permissible pressure for the piping.
The original SNUPPS submittal SLNRC 83-002 (1/7/83) states that all acoustic pressures in the upstream piping calculated or observed prior to and during safety valve hot or cold seal loop discharge are below the maximum permissible pressure.
In addition, the bending moments measured during the EPRI tests of Crosby valves with " reference" ring settings were
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all less than the maximum bending moment observed during test 908, i.e.
298,750 in-lb. Therefore, sufficient information exists to provide assur-c ance that no valve or piping system damage will occur as a result of these high frequency pressure oscillations.
LISTING OF REPORTS REFERENCED IN RESPONSES
- 1. EPRI Report NP-2296, Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valves in Westinghouse-Designed Plants, December 1982.
Test Condition Justification Report, December 1982.
Safety and Relief Valve Test Report, December 1982.
- 4. EPRI Report NP-2770-LD, Vol.
6., EPRI/CE PWR Safety Valve test Report:
Test Results for Crosby Safety Valve, March 1983.
- 5. WCAP-10105, Review of Pressurizer Safety Valve Performance as Observed in the EPRI Safety and Relief Valve Test Program, June 1982.
- 6. EPRI Summary Report: Westinghouse Gate Valve Closure Testing' Program, Engineering Memorandum 5683, Rev. 1, March 31, 1982.
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