ML20027D613
| ML20027D613 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 10/29/1982 |
| From: | Pilant J NEBRASKA PUBLIC POWER DISTRICT |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.A.1, TASK-TM NUDOCS 8211080159 | |
| Download: ML20027D613 (13) | |
Text
&~O9f h"~ kNebraska Public Power District GENERAL OFFICE
" " "7ssYs*E difa^e*st"* '"
October 29, 1982 U.S. Nuclear Regulatory Commission Attention:
Mr. Domenic B. Vassallo, Chief Operating Reactors Branch No. 2 Division of Licensing Washington, DC 20555
Reference:
Letter from D. B. Vassallo to J. M. Pilant dated July 13, 1982, Same Subject
Dear Mr. Vassallo:
Subject:
NUREG-0737 Item II.D.1, Relief and Safety Valve Test Requirements; Request for Additional Information Reference 1 requested the District to respond to six plant specific concerns arising from your review of NEDE-24988-P,
" Analysis of Generic BWR Safety / Relief Valve Operability Test Results. " Enclosed are the plant specific responses for Cooper Nuclear Station.
If additional clarification is necessary regarding the enclosed information, please do not hesitate to contact me.
Sincerely, M
a M. Pilant Division Manager of Licensing and Quality Assurance JMP/kew:bkn21/2 Enclosure eh g
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-NRC QUESTION 1 The test program utilized a." rams head" discharge pipe configuration._ Cooper Statibn utilizes a " tee"_ quencher configuration at the end of the discharge
- line.
Describe the discharge pipe' configuration used at Cooper Station and compare the anticipated loads on velve internals in the Cooper' Station con-
- figuration.to the measured loads in the test program.- Discuss the impact,of any differences in loads on valve operability.
-RESPONSE TO NRC' QUESTION 1-The safety / relief valve discharge piping configuration at Cooper Station utilizes'a " tee" quencher at the discharge pipe exit. The average length of the eight (8) SRV' discharge lines (SRVDL) is 140 feet and the submergence length in the suppression pool is approximately 12-feet. The SRV test prcgram utilized a rams head at the discharge-pipe exit, a pipe length of 112. feet and a. submergence length of approximately 13 feet.
Loads on valve internals
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during the-test-program are larger than loads ~on valve internals in the Cooper Station configuration for the following reasons:
1.
No dynamic mechanical load originating at the " tee" quencher is trans--
mitted to the valve in the Cooper Station configuration because there is at least one anchor point between the valve and the tee quencher.
~2.
The first length of the segment of piping downstream of the SRV in the test facility was longer than the Cooper Station piping, thereby result-ing in a bounding dynamic mechanical load on the valve in the test program. The first segment length in the test factlity is 12 ft whereas this length is 6 ft in the plact configuration.
3.
Dynamic hydraulic loads (backpressure)- are experienced by the valve internals in the Cooper Station' config'iration.
The backpressure loads may be either (i) transient backpressures occurring during valve actuation, or (ii) steady-state backpressures occurring during steady-state flow following valve actuation.:
(a) The key parameters affecting 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 backpressure in the test - program was maximized by utilizing a submergence of 13', which is greater than Cooper Station and a pipe length of _112' which is less' than Cooper Station. The maximum transient backpressure occurs with high pressure steam flow conditions. The transient backpressure for the alternate shutdown cooling mode of operation is always less than the design for steam ficw conditions because of the lower upstream pressure and the longer valve opening time.
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(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 rams head.
The orifice was sized to produce a back-pressure greater than that calculated for any of the Cooper Station SRVDL's.
The differences in the line configuration between Cooper Station and the test program as discussed above result in the loads on the valve internals for the test facility which bound the actual Cooper Station loads.
An additional consideration in the selection of the rams head for the test facility was to allow more direct measurement of the thrust load in the final pipe segment.
Utilization of a " tee" 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 at Cooper Station and the test facility will not heve any adverse effect on SRV operability at Cooper Station relative to the test facility.
NRC QUESTION 2 The test configuration utilized no spring hangers as pipe supports.
Plant specific configurations do use spring har.3ers in conjunction with snubber and rigid supports. Describe the safety relief valve pipe supports used at Cooper Station and compare the anticipated loads on valve internals for the Cooper Station pipe supports to the measured loads in the test program. Describe the impact of any differences in loads on valve operability.
RESPONSE TO NRC QUESTION 2 The Cooper Station safety-relief valve discharge lines (SEVDL's) are supported by a combination of snubbers, rigid supports, and spring hangers.
The locations of snubbers and rigid supports at Cooper Station are such that the location of such supports in the BWR generic test facility is prototypical, i.e.,
in each case (Cooper Station and the test facility) there are supports near each change of direction in the pipe routing. Additionally, SRVDL's at Cooper Station have between 1 and 5 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 considered generic to all BWR's since the test facility was designed to be prototypical of the featurcs pertinent to this issue.
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 offset 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.
This question addresses the design adequacy of the spring hangers with respect 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 from the high pressure steam discharge. Therefore, it-is believed that sufficient margin exists in the Cooper Station piping system design to adequately offset the increased dead load on the spring hangers in an unpinned condition due to a water filled condition.
Furthermore, the effect 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 1 cads occur in the SRVDL only after valve opening.
NRC QUESTION 3 Report NEDE-24988-P did not identify any valve functional deficiencies or anomalies during ' the test program.
Describe the impact on valve safety function of any valve functional deficiencies or anomalies encountered during the program.
RESPONSE TO NRC QUESTION 3 No functional deficiencies or anomalies of the safety relief or relief valves, were experienced during the testing at Wyle Laboracories for compliance with the alternate shutdown cooling mode requirement. All 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 i
due to failures of test facility instrumentation, equipment, data acquisition
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equipment, or deviation from the approved test procedure.
The test specification for each valve required six runs.
Under the test ju'ged invalid.
All.
procedure, any anomaly caused the test run to be d
anomalies were reported in the test report.
The Wyle Laboratories test log sheet for the two-stage target rock valve tests is attached.
This valve is used at Cooper Nuclear Station.
Each Wyle test report for the respective valves identifies each test run performed and documents whether or not the test run is 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.
function.
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:
4 (a) Presenting the maximum representative loading information obtained from the steam run data, (b) Presenting the maximum representative water loading information obtained from the 15'F subcooled water test data,
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(c) Presenting the data on the only test run performed for the 50*F subcooled water test condition.
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 Cooper. Station 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 plant features at Cooper Station.
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 Cooper Station.
RESPONSE TO NRC QUESTION 4 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 the dynamic forces on the safety and relief valves would be maximized. Test pressures were the highest predicted by conventional safety analysis procedures.
The BWR Owners Group, in their enclosure to the September 17, 1980, letter from 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 dxpected 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 Cooper Station 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 program, as documented in NEDE-24988-P, are 15*F to 50*F subcooled liquid at 20 psig to 250 psig. These fluid conditions envelope the conditions expected to occur at Cooper Station 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 component failure or operator error postulated in the events sequence.
These events and the plant-specific features that
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mitigate these events, are summarized in Table 1.
Of these 13 events, only ten (10) are' applicable to Cooper because of its design and specific plant configuration.
Thr,+ events, namely 5, 6, and 10 are not applicable to Cooper Station for to ressais listed below:
a.
Event Number
- fransient High Pressure Core Spray This system is not installed at Cooper.
b.
Event Number 6 - Transient RCIC Head Spray This system is not installed at Cooper.
c.
Event Number 10 - Small Break Accident, High Pressure Core Spray This system is not installed at Cooper.
Of these ten (10) remaining events, the Cooper Station 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 Cooper Station because the base case analysis does not include any plant features which are not already present 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 Cooper Station plant. All features included in this base case analysis are similar to plant features in the Cooper Station design.
Furthermore, the time available for operator action, is expected to be longer in the Cooper Station plant than in the base case analysis for each case where operator action is required.
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.
At Cooper Station, this event involves flow of subcooled water (approximately 44*F subcooled) at a pressure of approximately 75 psig.
The test conditions clearly envelope these plant conditions.
As discussed above, the BWR Owners Group evaluated transients including single active failures that would maximize the dynamic forces on the safety relief valves. 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 Cooper Station 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 performance 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 Cooper Station. 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 specificatien limit of 100*F per hour.
When the vessel is depressurized, the operator initiates normal shutdown cooling by use of the RHR system.
If that system is unavailable because the valve on the RHR shutdown cooling suction line fails to open, the operator initiates the alternate shutdown cooling mode.
For alternate shutdown cooling, the operator opens one SRV and initiates either an RHR or core spray pump utilizing the suppression pool as the suction source. The reactor vessel is filled such that water is allowed to flow into the main stream lines and out of the SRV and back to the suppression pool.
Cooling of the system is provided by use of an RHR 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 performcd.
In order to control the reactor vessel cooldown rate, the operator would 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 Cooper Station SRV to be extensively cycled for steam discharge conditions has been confirmed during steam discharge qualification testing of the valve by the valve vendor. Based on the qualification testing I
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 l
the probability of the valve to fail open or closed is extremely low.
I NRC QUESTION 6 l
Describe how the values of Valve C 's in report NEDE-24988-P will be used at Cooper Station.
Show that the melhodology used in the test program to determine the valve C will be consistent with the application at Cooper i
Station.
l _
RESPONSE TO NRC QUESTION 6 The flow coefficient, C, for the Target. Rock safety relief valve (SRV) y (2 stage) utilized at Cooper _ Station was determined in the generic SRV test program (NEDE-24988-P).
The average flow coefficient calculated from the test results for the Target Rocks, is reported in Table 5.2-1 of NEDE-24988-P.
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This test value has been used by BWR plants to confirm that the liquid discharge flow capacity of the SRV's will be sufficient to remove core decay-heat when injecting into the reactor pressure vessel (RPV) in the alternate shutdown cooling mode.
CNS does not rely upon the. alternate shutdown cooling mode as the licensing basis in the FSAR. The Cv.value determined in the
. SRV test indicates that the Cooper Station SRV's are capable of returning the flow injected _by.the RHR or CS pump'to_the suppression pool.
If it were necessary for the operator to place Cooper Station in the alternate shutdown cooling mode, he would assure that adequate core cooling was being provided by monitoring the following parameters: LRHR or CS flow rate, reactor i
vessel. pressure and reactor vessel temperature.
The flow coefficient for the Target Rock 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 valve inlet (steam chest) and 3' downstream of the valve and the corresponding measured flowrate.
Furthermore, the test conditions and test configuration: vere judged representative of Cooper Station plant conditions for the alternate shutdown cooling mode, e.g., pressure upstream of the valve, fluid temperature, friction losses and liquid flowrate, although the alternate shutdown cooling mode is not required to be fully analyzed for CNS. 'Therefore, the reported-Cy values-are appropriate for application to Cooper Station.
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TEST LOG FOR SRV TR-1 I
Test Test Load Line Test q
No.
Media Configuration Date Remarks ti
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