ML20204G132
| ML20204G132 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 04/27/1983 |
| From: | Fiedler P GENERAL PUBLIC UTILITIES CORP. |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM NUDOCS 8305020362 | |
| Download: ML20204G132 (21) | |
Text
,
s GPU Nuclear NUQ Mf P.O. Box 388 Forked River, New Jersey 08731 609-693-6000 Writer's Direct Dial Number April 27, 1983 Sy -zL c3 Mr. Dennis M. Crutchfield, Chief Operating Reactors Branch #5 U.S. Nuclear Regulatory Commission Wa shing ton, D.C.
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Dear Mr. Crutchfield:
l l
Subj ect: Oyster Creek Nuclear Generating Station l
NUREG - 0737 Item II.D.I Safety / Relief Valve Testing I
l As requested by your correspondence of December 17,1982, we are submitting our plant specific responses to your six questions on Safety / Relief Valve Testing. Should you have any further questions on this subject please contact Mr. James Knubel, Manager BWR Licensing at l
(201) 299-2264.
l Very truly yours, l
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Peter B. Fiedler Vice President and Director Oyster Creek PBF:jal Enclosure cc: Regional Administrator hOl Region I U.S. Nuclear Regulatory Commission I
631 Park Avenue King of Prussia, PA 19406 NRC Resident Inspector Oyster Creek Nuclear Generating Station Forked River, NJ 08731 8305020362 830427 PDR ADOCK 05000219 P
PDR GPU Nuclear is a part of the General Public Utilities System
4 NRC QUESTION 1 The test program utilized a " rams head" discharge pipe configuration. Oyster Creek utilizes a "Y" quencher configuration at the end of the discharge line.
Describe the discharge pipe configuration used at Oyster Creek and compare the anticipated loads on valve internals in the Oyster Creek configuration to the i
measured loads in the test program. Discuss the imoset of any differences in loads on valve operability.
RESPONSE TO QUESTION 1 The safety / relief valve discharge piping configuration at Oyster Creek utilizes a "Y" quencher with a sparger at the discharge pipe exit. The minumum length of the two (2) SRV discharge lines (SRVDL) is 103 f t. the maximum is 116 ft.
The submergence length in the suppression pool is approximately 8 feet. The SRV test program utilized a ramshead at the discharge pipe exit, a pipe length of 112' and submergence length of approximately 13'. Loads on valve internals during the test program are larger than loads on valve internals in the Oyster Creek configuration for the following reasons:
1.
No dynamic mechanical load originating at the "Y" quencher is transmitted to the valve in the Oyster Creek configuration because there is at least.one anchor point between the valve and the "Y" quencher.
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2.
The first length of the segment of piping downstream of the SRV in the test facility was longer than the Oyster Creek piping, thereby resulting in a bounding dynamic mechanical load on the valve in the test program due to the larger moment arm between the SRV and the first elbow. The first segment length in the test facility is 12 f t.
whereas this length is 2 f t. in the plant configuration.
3.
Dynamic hydraulic loads (backpressure) are experienced by the valve internals in the Oyster Creek configuration. The backpressure loads may be either (1) transient backpressures occurring during valve actuation, or (ii) steady-state backpressures occurring during steady-state flow following valve actuation.
(a) The key parameters af fecting the transient backpressures are the fluid pressure upstream of the valve, the valve opening time, the fluid inertia in the submerged SRVDL and the SRVDL air volume. Transient backpressures increase with higher upstream pressure, shorter valve opening times, greater line submergence, and smaller SRVDL air volume. The transient back pressure in the test program was maximized by utilizing a submergence of i
13', which is greater than Oyster Creek and a pipe length of 3
112' (10" sch 80: volume = 96,513 in ) which is less than Oyster 3
Creek, (minimum volume is 124,000 in ; maximum volume is 3
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s The maximum transient backpressure occurs with high pressure steam flow conditions. The transient backpressure for the alternate shutdown cooling mode of operation is always much less than the design for steam flow conditions because of the lower upstream pressure and the longer valve opening time.
(b) The steady-state backpressure in the test program was maximized by utilizing an orifice plate in the SRVDL above the water level and before the ramshead. The orifice was sized to produce a backpressure greater than that calculated for any of the Oyster Creek, SRVDL's.
The dif ferences in the line configuration between the Oyster Creek plant and the test program as discussed above result in the loads on the valve internals for the test facility which bound the actual Oyster Creek loads. An additional consideration in the selection of the ramshead for the test facility was to allow more direct measurement of the thrust load in the final pipe segment.
Utilization of a "Y" quencher in the test program would have required quencher supports that would unnecessarily obscure accurate measurement of the pipe thrust loads. For the reasons stated above, differences between the SRVDL configurations in Oyster Creek and the test facility will not have any adverse ef fect on SRV operability at Oyster Creek relative to the test facility.
s NRC QUESTION 2 The test configuration utilized no spring hangers as pipe supports. Plant specific configurations do use spring hangers in conjuction with snubber and rigid supports. Describe the safety relief valve pipe supports used at Oyster Creek and compare the anticipated loads on valve internals for the Oyster Creek pipe supports to the measured loads in the test program. Describe the impact of any dif ferences in loads on valve operability.
I RESPONSE TO QUESTION 2 The Oyster Creek' safety-relief valve discharge lines (SRVDL's) are supported by a combination of snubber, rigid supports, and spring hangers. The locations of snubbers and rigid supports at Oyster Creek are such that the location of such supports in the BWR generic test facility is prototypical, i.e., in each case (Oyster Creek and the test facility) there are supports near each change of direction in the pipe routing. Additionally, each SRVDL at Oyster Creek has only 2 spring hangers, all of which are located in the drywell. The spring hangers, snubbers, and rigid supports were designed to accommodate combinations of loads resulting from piping dead weight, thermal conditions, seismic and suppression pool hydrodynamic events, and a high pressure steam discharge transient.
The dynamic load effects on the piping and supports of the test facility due to the water discharge event (the alternate shutdown cooling mode) were found to be significantly lower than corresponding loads resulting from the high pressure steam discharge event. As stated in NEDE-24988-P, this finding is
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l considered generic to all BWR's since the test facility was designed' to be I
prototypical ~ of the features pertinent to this issue. Furthermore, analysis of
- a typical Oyster, Creek SRVDL configuration has confirmed the applicability of this conclusion to Oyster Creek.
During the water discharge transient there will be significantly lower dynamic loads acting on the snubbers and rigid supports than during the steam discharge transient. This will more than of fset the small increase in the dead load on -
these supports due to the weight of the-water during the alternate shutdown cooling mode of operation. Therefore, design adequacy of the snubbers and '
rigid supports is assured as they are designed -for the larger steam discharge transient loads.
I This question addresses the design adequacy of the spring hangers with respect i'
i to the increased dead load due to the weight of the water during the liquid
' discharge transient. As was discussed with respect to snubbers and rigid supports, the dynamic. loads resulting from liquid discharge during the alternate shutdown cooling mode of operation are significantly lower than those i
from the high pressure steam discharge. Therefore, it is believed that l
suf ficient margin exists in the Oyster Creek piping system design to adequately i
of fset the increased dead load on the spring hangers in an unpinned condition due to a water filled condition. Furthermore, the ef fect of the water dead weight load does not affect the ability of SRV's to open to establish the alternate shutdown cooling path since the loads occur in the SRVDL only af ter l
valve opening.
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OPERABILITY TEST REPORT mR DRESSER 6x8 SRV LOW PRESSURE WATER TESTS FOR GENERAL ELECTRIC COMPANY
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NRC QUESTION 3 Report NEDE-24988-P did not identify any valve functional deficiencies or anomalies encountered during the test program. Describe the impact on valve safety function of any valve functional deficiencies or anomalies encountered during the program.
RESPONSE TO QUESTION 3 No functional deficiencies or anomalies of the safety relief or relief valves were experienced during the testing at Wyle Laboratories for compliance with the alternate shutdown cooling mode requirement. All of the valves subjected to test runs, valid and invalid, opened and closed without loss of pressure integrity or damage. Anomalies encountered during the test program were all due to failures of test facility instrumentation, equipment, data acquisition equipment, or deviation from the approved test procedure.
The test specification for each valve required six runs. Under the test procedure, any anomaly caused the test run to be judged invalid. All anomalies were reported in the test report. The Wyle Laboratories test log sheet for the Dresser Electromatic 1525-VX valve tests is attached. This valve is used in the Oyster Creek Nuclear Power Station.
Each Wyle test report for the respective valves identifies each test run performed and documents whether or not the test run iv valid or invalid and states the reason for considering the run invalid.
No anomaly encountered during the required test program affects any valve safety or operability func tion.
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All valid test runs are identified in Table 2.2-1 of NEDE-24988-P. The data presented in Table 4.2-1 for each valve were obtained from the Table 2.2-1 test runs and were based upon the selection criteria of:
(a) Presenting the maximum representative loading information obtained from the steam run data, (b) Presenting the maximum representative water loading information obtained f rom the 15 F subcooled water test data, i
(c) Presenting the data on the only test run performed for the 50 F subcooled water test condition.
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s TEST. REPORT NO. 17476-06 TABLE 1 OPERABILITIY TEST LOG, SRV DR-1 TEST LOAD LINE TEST No.
MEDIA CONFIGURATION DATE REMARKS 601 Steam 1
4/15/81 Back pressure too high.
602 Steam 1
4/15/81 Installed 5.75" orifice.
Test acceptable.
603 Water 1
4/15/81 Steam chest pressure low.
604 Wa ter 1
4/15/81 Test acceptable.
605 Steam 1
4/16/81 No data on tape.
606 Wa ter 1
4/16/81 Test acceptable.
607 Steam 1
4/16/81 Test acceptable.
608 Wa ter 1
4/16/814 Test acceptable.
i 609 Steam 1
4/16/81 Rerun of Test #605.
l Test acceptable.
Replaced L1 snubber for 608 and 609.
l l
WYLE LABORATORIES 3
l Huntsville Facility l
l l
NRC QUESTION 4 The purpose of the test program was to determine valve performance under conditions anticipated to be encountered in the plants. Describe the events and anticipated conditions at Oyster Creek for which the valves are required to operate and compare these plant conditions to the conditions in the test program. Describe the plant features assumed in the event evaluations used to scope the test program and compare them to the plant features at Oyster Creek.
For example, describe high level trips to prevent water from entering the steam lines under high pressure operating conditions as assumed in the test event and compare them to trips used at Oyster Creek.
RESPONSE TO NRC QUESTION 4 I
The purpose of the S/RV test program was to demonstrate that the Safety Relief Valve (S/RV) will open and reclose under all expected flow conditions. The expected valve operating conditions were determined through the use of analyses of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70, Revision 2.
Single failures were applied to these analyses so that i
the dynamic forces on the safety and relief valves would be maximized. Test pressures were the highest predicted by conventional safety analysis pro cedure s.
The BWR Owners Group, in their enclosure to the September 17, 1980 letter f rom D.B. Waters to R.H. Vollmer, identified 13 events which may result in liquid or two-phase S/RV inlet flow that would maximize the dynamic forces on the safety and relief valve. These events were identified by evaluating the
initial events described in Regulatory Guide 1.70, Revision 2, with and without the additional conservatism of a single active component failure or operator error postulated in the event sequence.
It was concluded from this evaluation that the alternate shutdown cooling mode is the only expected event which will result in liquid at the valve inlet. Consequently, this was the event simulated in the S/RV test program. This conclusion and the test results applicable to Oyster Creek are discussed below. - The alternate shutdown cooling mode of operation has been described in the response to NRC Question 5.
The S/RV inlet fluid conditions tested in the BWR Owners Group S/RV test 0 to 500 subcooled liquid at 20
' program, as documented in NEDE-24988-P, are 15 psig to 250 psig. These fluid conditions envelope the conditions expected to occur at Oyster Creek in the alternate shutdown cooling mode of operation.
The BWR Owners Group identified 13 events by evaluating the initiating events described in Regulatory Guide 1.70, Revision 2, with the additional conservatism of a single active componet failure or operator error postulated in the events sequence. These events and the plant-specific features that mitigate these events, are summarized in Table 1.
Of these 13 events, only 6 are applicable to the Oyster Creek plant because of its design and specific i
plant configuration. Seven (7) events, namely 3,4,5,6,9,10, and 11 are not applicable to the Oyster Creek plant for the reasons listed below:
(a) Events 5 and 10 are not applicable because Oyster Creek does not have an HPCS system.
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(b) Events 3 and 11 are not applicable because Oyster Creek does not have an HPCI system.
(c) Events 4, 6, and 9 are not applicable because Oyster Creek does not have an RCIC system.
J For the 6 remaining events, the Oyster Creek specific features, such as trip logic, power supplies,: instrument line configuration, alarms and operator actions, have been compared to the base case analysis presented in the BWR Owners Group submittal of September 17, 1980. The comparison has demonstrated that in each case, the base case analysis is applicable to Oyster Creek because the base case analysis does not include any plant features which are not already.
present in the Oyster Creek design.
For these events, Table 1 demonstrates that the Oyster Creek specific features are included in the base case analyses presented in the BWR Owners Group submittal of September 17, 1980. It is seen from Table 1, that all plant features assumed in the event evaluation are also existing features in the Oyster Creek plant. All features included in this base j
case analysis are similar to plant features in the Oyster Creek design.
Furthermore, the time available for operator action is expected to be longer in the Oyster Creek plant than in the base case analysis for each case where operator action is required.
l Event 7, the alternate shutdown cooling mode of operation, is the only expected event which will result in liquid or two-phase fluid at the S/RV inlet.
Consequently, this event was simulated in the BWR S/RV test program. In Oyster 4
Creek, this event involves flow of subcooled water (up to 500F subcooled) at a pressure of approximately 150 psig. The test conditions clearly envelope these i
plant conditions.
i.
As discussed above, the BWR Owners Group evaluated transients including single active failures that would maximize the dynamic forces on the safety relief valve s.
As a result of this evaluation, the alternate shutdown cooling mode is the only expected event involving liquid or two-phase flow. Consequently this event was tested in the BWR S/RV test program. The fluid conditions and flow conditions tested in the BWR Owners Group test program conservatively envelope the Oyster Creek plant specific fluid conditions expected for the alternate shutdown cooling mode of operation.
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NRC QUESTION 5 The valves are likely to be extensively cycled in a controlled depressurization mode in a plant-specific -application. Was this mode simulated in the test program? What is the effect of this valve cycling on valve performtnce and probability of the valve to fail open or to fail closed?
RESPONSE TO NRC QUESTION 5 The BWR safety / relief valve (SRV) operability test program was designed to simulate the alternate shutdown cooling mode, which is the only expected liquid discharge event for Ofster Creek. The sequence of events leading to the alternate shutdown cooling mode is given below.
Following normal reactor shutdown, the reactor operator depressurizes the reactor vessel by opening the turbine bypass valves and removing heat through the main condenser. If the main condenser is unavailable, the operator could depressurize the reactor vessel by using the SRV's to discharge steam to the suppression pool. If SRV operation is required, the operator cycles the valves in order to assure that the cooldown rate is maintained within the technical specification limit of 1000F per hour. When the vessel is depressurized, the operator initiates normal shutdown cooling by use of the Shutdown Cooling System. If that system is unavailable because the valve on the Shutdown Cooling System shutdown cooling suction line fails to open or the Reactor
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Building Closed Cooling Water System is not operable, the operator initiates the alternate shutdown cooling mode.
For alternate shutdown cooling, the operator opens one SRV and initiates one core spray system utilizing the suppression pool as the suction source. The reactor vessel is filled such that water is allowed to flow into the main tream lines and out of the SRV and back to the suppression pool. Cooling of the system is provided by the use of a containment spray heat exchanger. - As a result, an alternate cooling mode is maintained.
In order to assure continuous long term heat removal, the SRV is kept open and no cycling of the valve is performed. In order to control the reactor vessel cooldown rate, the operator is instructed to control the flow rate into the vessel. Consequently, no cycling of the SRV is required for the alternate shutdown cooling mode, and no cycling of the SRV was performed for the generic BWR SRV operability test program.
The ability of the Oyster Creek SRV to be extensively cycled for steam discharge conditions has been confirmed during steam discharge testing of the valve by the valve vendor. Based on the testing of the SRV's, the cycling of the valves in a controlled depressurization mode for steam discharge conditions will not adversely af fect valve performance and the probability of the valve to fail open' or closed is extremely low.
NRC QUESTION 6 Describe how the values of valve Cy 's in report NEDE-24988-P will be used at Oyster Creek.
Show that the methodology used in the test program to determine the valve C will be consistent with the application at Oyster Creek.
y RESPONSE TO NRC QUESTION 6 The flow coef ficient, Cy, for the Dresser Electromatic 1525-VX safety relief valve (SRV) utilized in Oyster Creek was determined in the generic SRV test program (NEDE-24988-P). The average flow coefficient calculated from the test results for the Dresser 1525-VX, is reported in Table 5.2-1 of NEDE-24988-P.
This test value has been used by GPU Nuclear to confirm that the liquid discharge flow capacity of the Oyster Creek SRV's will be suf ficient to remove core decay heat when injecting into the reactor pressure vessel (RPV) in the alternate shutdown cooling mode.- The Cy value determined in the SRV test demonstrates that the Oyster Creek SRV's are capable of returning the flow injected by a CS pump to the suppression pool.
If it were necessary for the operator to place the Oyster Creek plant in the alternate shutdown cooling mode, he would assure that adequate core cooling was being provided by monitoring the following parameters:
CS flow rate, reactor vessel pressure and reactor vessel temperature.
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The flow coefficient for the Dresser Electromatic 1525-VX valve reported in NEDE-24988-P was determined from the SRV flow rate when the valve inlet was pressurized to approximately 250 psig. The valve flow rate was measured with the supply line flow venturi upstream of the steam chest. The Cy for the valve was calculated using the nominal measured pressure differential between the 4
valve inlet (steam chest) and 3' downstream of the valve and the corresponding measured flowrate. Furthermore, the test conditions and test confirguration were representative of Oyster Creek plant conditions for the alternate shutdown t
cooling mode, e.g. pressure upstream of the valve, fluid temperature, friction losses and liquid flowrate. Therefore, the reported Cy values are appropriate for application to the Oyster Creek plant.
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