ML20064K237
| ML20064K237 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 01/14/1983 |
| From: | Zimmerman S CAROLINA POWER & LIGHT CO. |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM 5876C10T2, NUDOCS 8301180337 | |
| Download: ML20064K237 (18) | |
Text
CD&L Carohna Power' & Light Company JAN 141933 Of fice of Nuclear Reactor Regulation ATTN:
Mr.
D. B. Vassallo, Chief Operating Reactors Branch No. 2 United States Nuclear Regulatory Commissica Washington, DC 20555 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324 LICENSE NOS. DPR-71 AND DPR-62 NUREG-0737 ITEM I1.D.1 RELIEF AND SAFETY VALVE TESTING
Dear Mr. Vassallo:
In response to your letter of July 12, 1982 concerning NUREG-0737 Item II.D.1, Relief and Safety Valve (S/RV) Testing, Carolina Power & Light Company (CP&L) is providing responses to your questions concerning the final results on S/RV testing for the Brunswick Steam Electric Plant, Unit Nos. I and 2.
The final test results are contained in NEDE-24988-P, " Analysis of Generic BWR Safety / Relief Valve Operability Test Results," which were submitted to the NRC by the BWR Owners' Group on September 25, 1981.
The applicability of the generic test results to the Brunswick S/RV's was demonstrated in our June 30, 1981 and Oc tober 5, 1981 submittals.
If you have any questions concerning this subject, please contact our staff.
Yours very truly, f
l')2/ 7P t*: n S.
mmerman Manager Licensing & Permits WRM/kjr (5876C10T2)
Enc losure cc:
Mr. S. D. MacKay (NRC)
Mr. D. O. Myprs (NRC-BSEP) 4[
Mr. J.
P. O'Reilly (NRC-RII)
O iAs Mr. J. A. Van Vliet (NRC)
I 8301180337 830114 PDR ADOCK 05000324 p
PDR 411 Fayetteville Street
- P. O. Box 1551
- Raleigh, N. C. 27602
e 1
i NRC QUESTION NO. 1:
The test program utilized a " rams head" discharge pipe configuration.
Brunswick utilizes a " tee" quencher configuration at the end of the discharge line.
Describe the discharge pipe configuration used at Brunswick and compare the anticipated loads on valve internals in the plant configuration to the measured loads in the test program.
Discuss the impact of any differences in loads on valve operability.
CP&L Response:
The safety / relief valve (S/RV) discharge piping configuration at Brunswick utilizes a " tee" quencher at the discharge pipe exit.
The average length of the eleven S/RV discharge lines (SRVDL) is 75 feet and the submergence length in the suppression pool is approximately 7 feet.
The S/RV test program utilized a ramshead at the discharge pipe exit, a pipe length of 112 feet and a submergence length of approximately 13 feet.
Loads on valve internals during the test program are larger than loads on valve internals in the
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-ick configuration for the following reasons:
do dynamic mechanical load originating at the " tee" quencher. is transmitted to the valve in the Brunswick 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 S/RV in the test f acility was longer than the 3runswick piping, thereby resulting in a bounding dynamic mechanical load on the valve in the test program due to the larger moment arm between the S/RV and the first elbow.
The first segment length in the test facility is 12 feet whereas these lengths are 1.5 to 12 feet in the plant configuration.
3.
Dynamic hydraulic loads (backpressure) are experienced by the valve internals in the Brunswick configuration. The backpressure loads may be either (1) transient backpressures occurring during valve actuation, or (ii) steatr-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 che 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 feet, which is greater than Brunswick, and a pipe length of 112 f eet, which is more than Brunswick.
the maximum transient backpressure occurs with high pressure steam flos 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 naximized by utilizing a 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 Brunswick SRVDL's.
The differences in the line configuration between the Brunswick plant and the test program as discussed above result in the loads on the valve internals of the test facility which bound the actual Brunswick 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
" 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 SRDVL configurations in Brunswick and the test facility will not have any adverse ef fects on S/RV operability at Brunswick relative to the-test f ac ility.
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NRC OUESTION No. 2:
Plant The test configuration utilized no spring hangers as pipe supports.
specific configurations do use spring hangers in conjunction with snubber and Describe the safety relief valve pipe supports used at rigid supports.
Brunswick and compare the anticipated loads on valve internals for the plant Describe the impac t pipe supports to the measured loads in the test program.
of any dif ferences in loads on valve operability.
The Brunswick safety / relief valve discharge lines (SRVDL's)
CP&L Respons,e:
are supported by a combination of snubbers, rigid supports, and spring hangers. The locations of snubbers and rigid supports at 3runswick are such that the location of such supports in the BWR generic test f acility is prototypical, i.e., in each case (Brunswick and the test facility) there are supports near each change of direction in the pipe Additionally, each SRVDL at Brunswick has only 2,3, routing.
or 4 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 hydrodynanic events, and a high pressure steam discharge transient.
The dynamic load ef fects on the piping and supports of the f acility due to the water discharge event (the alternate test shutdown cooling mode) were f ound to be significantly loser than corresponding loads resulting from the high pressure steam discharge event.
As s tated in NEDE-24988-P, this finding is considered generic to all mid's.since the test f acility was designed to be prototypical of the features to this issue.
Futhermore, analysis of a typical pertinent Brunswick SRVDL configuration has confirmed the applicability of this conclusion to Brunswick.
During the water discharge transient there will be significantly lower dynamic loads acting on the snubbers and rigid supports than during the s team discharge t ransient.
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 addcesses the design adequacy of the spriag 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 ric;id supports, the dynamic loads resulting f rom liquid discharge durf ng the alternate shutdosa cooling node of operation are significantly lower than those f rom the high pressure steam
discharge.
Hierefore, it is believed that suf ficient mardin exists in the Brunswick piping systen design to adequately of fset the increased dead load on the spring hangers in an unpinned condition due to a water filled condition.
Futhermore, the effect of the water dead weight load does not affect the ability of safety / relief valves (S/RV's) to open to establish the alternate shutdown cooling path since the loads occur in the SRVDL only af ter valve opening.
NRC QUESTION NO. 3:
Feport NEDE-24988-P did not identify any valve functional deficiencies or anomalies encountered during the test program.
Describe the impact of valve safety function of any valve functional deficiencies or anomalies encountered during the program.
CP5L Response:
No functional defic'.encies or anomalies of the safety relief or relief valves we re experienced during the testing at Wyle Laboratories for compliance with the alternate shutdown cooling mode requirement.
All of the valves subjected to tett runs, valid and invalid, opened and closed without loss f pressure integrity or damage.
Anomalies encountered during the test program were all due to failures of test facility instrumentation, equipment, data acquirition equipment, or deviation f rom 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 Target Rock two-stage valve tests is attached.
This valve is used in the Brunswick Plant.
Each Wyle test report for the respective valves identifies each test run performed, 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 af fects any valve safety or operability f unc tion.
All valid test runs are identified in Table 2,2-1 of NEDE-2 '4988-P.
The data presented in Table 4.2-1 ( Attachment
- 1) for each valv 2 were obtained f rom 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 maximur representative water loading information obtained t.om the 15'F subcooled water test
- data, (c )
Presenting the data on the only test run performed f or the 50*F subcooled water test condit.on.
ATTACfDIENT I (Fesponse No. 3)
NEDE-24988-P ANALYSIS OF GENERIC BWR SAFETY / RELIEF VALVE OPERABILITY TEST FISULTS
CABLE 4.2-1 SU.3:4ARY OF REGUCED DATA TARGET ROCK 6X10-2 STAGE S/RV WITil LOADS I SUPPORTS Test Data
- Steam, Water, 15*F Water, 50*F Test Parame te r Saturated Subcooling Subcooling Desc rip tion Uitits Run 301 Rutt 303 Run 307 Fluid inlet temperature
- F 555 249 215 Steam ilow rate 838,900 N/A N/A at (
) psig lbs/hr (1128)
Average backpressure (Wyle Data) 25.5 N/A N/A Operability, open/ closed upon command yes/no yes yes yes Opening time, main valve disc ms ec 24 Average water pressure psig N/A 270 264 Average water flow rate gpm N/A 6784 6619 S/RV test facility integrity, yes/no yes yes yes after run 3/RV integrity, post test hydro yes/no yes yes yes S/RV internal parts lategrity, post test disassembly /inspec tion yes/no N/A N/A yes l
36
TABLE 4.2-1 SU:LiARY OF ltEDUCED DATA TARGET ROCK 6X10-2 STAGE S/RV WIT 11 LOADS I GUPPORTS Test Data. Maximum Dynamic Values
- Steam, Water, 15'F Water, 50 M-Test Parameter Saturated Subcooling Subcooling uesc rip tion Units Run 301 Run 303 Run 307 SRVDL Acceleration ( A2) - 2nd Sec tion "g's" 2
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<1 Load Cell ( Ll/, let lloriz. Support LBS 2,000 1,000 1,000 Load Cell (L3), 2nd lioriz. Support LBS 13,000 3,000 2,000 Load Cell (L2), 3rd lioriz. Support LBS 12,000 5,000 4,000 Load Cell (L4), 4th Ibriz. Suppoc*
LdS 26,000 2,000 2,000 St ress (SG9), 5th Horiz. Support psi 3,500 100 200 Stress (SG10), Steam Chest psi 700 500 200 Stress (SGil), Steam Chest - Middle psi 1,000 500 200 Stress (SG12), Steam Chest psi 700 500 200 St ress (SG13), Sweepolet psi 800 600 500 Stress (SG14), Sweapolet psi 2,000 600 500 St ress (SG15), Sweepolet psi 700 500 500 Stress (SG16), Saeepoiet psi 2,000 700 1,300 37
TABLE 4. 2-1 SUMMRY OF REDUCED DATA TARGET ROCK 6X10-2 STAGE S/RV WITil LOADS I SUPPORTS Test Data, Maximu:n Dyna:nic Values '
- Steam, Water, 15'F Water, 50*F Test Parameter Saturated Subcooling Subcooling Desc rip tion Units Run 301 Run 303 Run 307 SRVUL Stress (SG21), 1st Section psi' 200 200 200 SRVDL Stress (SG22), 1st Section psi 1,500 500 500 SRVDL Stress (SG23), 1st Section psi 200 100 200 SRVDL Stress (SG24), 1st Section psi 1,500 400 500 SRVDL Stress (SG25), 2nd Section psi 700 200 200-SRVDL Stress (SG26), 2nd Section psi 800 200 200 SRVDL Stress (SG27), 3rd Section psi 700 50 200 SRVDL Stress (SG28), 3rd Section psi 700 100 200 SRVDL Stress (SG29), 3rd Section psi 700 100 200 SRVDL Stress (SG30), 3rd Section psi 700 50 200 SRVDL Stress (SGS), 4th Section psi 1,000 100 200 SRVDL Stress (SG6), 4th Section psi 1,000 109 200 SRVDL Stress (SG7), 4th Section psi 1,500 100 200 SRVDL Stress (SG8), 4th Section psi 3,200 100 200 e
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HRC CUESTION HO. 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 in the plants. Describe the events and anticipated conditions at Brunswick 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 Brunswick. 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 l
them to trips used at Brunswick.
CP&L Resoonse:
The purpose of the safety / relief valve (S/RV) test program was to demonstrate that the S/RV vill open and reclose under
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all expected flow conditions. The expected valve operating conditions were determined through the use of analyses of accidents and anticipated operational occurrences referenced i
in Regulatory Guide 1.70, Revision 2.
Single failures were applied to these analyses so that the dynamic forces on the safety and re_ief valves would be maximised. Test pressures were the highest predicted by conventional safsty 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 uould maximise 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 i
from this evaluation that the alternate shutdown cooling code j
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 Brunswick are discussed below. The aloernate shutdown cooling mode of operation has been described in the response to NRC Question No. 5.
i The S/RV inlet fluid conditions tested in the BWR Owners' Group S/RV test program, as documented in NEDE-24988-P, are 15, to 50, subcooled liquid at 20 psig to 250 psig. These fluid conditions envelope the conditions expected to occur at Brunswick in the alternate shutdown cooling mode of j
operation.
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The BWR Owners' Group identified 13 events by evaluating the initiating events acecribed in Regulatory Guide 1.70, Revision 2, uith the additional conservatism of a single active component failure or operator error postulated in the i
events sequence. These events and the nlant-scecific I
features that mitigate those events, are summarisco in Table 1.
Af these 13 events. rnly 10 are applicaole to :na i
Brunswick plant because of its cesign anc apecific plant i
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For the 10 remaining events, the Brunswick specific features, such is trip logic, power supplies, instrument line configuration, alarms and operator actions, have been compared to the bcse case analysis presented in the BWR Owners' Group submittal of September 17, 1980. The comparison has demonstrated that in each case, the base case
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analysis is applicable to Brunswick because the base case analysis does not include any plant features which are not already present in the Brunswick design or affect the transient. For these events, Table 1 demonstrates that the Brunswick 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 (Attachment 2),
that all plant features assumed in the event evaluation are also existing features in the Brunswick plant. All features included in this base case analysis are similar to plant 4
features in the Brunswick design.
Furthermore, the time available for operator action is expected to be longer in the Brunswick plant than in the base case analysis for each case where operator action is required.
4 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.
The test conditions envelope the plant conditions and will be bounded by the Emergency Procedures.
As discussed above, the BWR Owners' Group evaluated transients including single active failures that would maximize the dynamic forces on the S/RVs.
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 Brunswick plant-specific fluid conditions expected for i
the alternate shutdown cooling mode of operation.
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TABLE 1 - EVENTS EVALUATED NOTES 1.
Reactor will scram if at greater than 30% power. Below 30% power, reactor scram would occur due to reactor low water level after reactor feed pumps are tripped or by manual scram.
2.
Reactor scram from MSIV closure only occurs in the Run mode. Reactor scram would occur from reactor low water level after reactor feed pumps are tripped in the Startup mode.
3 No HPCS System at Brunswick Plant.
4.
No head spray to RCIC Systs at Brunswick Plant.
5.
Only if reactor pressure is less than 410 psi. Low pressure would exist for a LBA.
6.
MSIV closure on low turbine inlet pressure occurs only in the Run mode.
In tne Startup mode, MSIV closure would occur after the reactor feed pumps trip due to reactor low water level or other existing signal.
7.
While RCIC would not automatically initiate on high drywell pressure, it would initiate on the reactor low water level that would exist during a LBA.
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NRC OUCSTION NO. 3:
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 close?
CP&L Response:
The BWR safety / relief valve (S/RV) operability test program was designed to simulate the alternate shutdown cooling mode, which is the only expected liquid discharge event for Brunswic k.
Following normal reactor shutdown, the reactor operator depressurises 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 S/RV's to discharge steam to the suppression pool.
If S/RV operation is required, the operator cycles the valves in order to assure that the cooldown rate is maintained within the technical specification limit of 100*F per hour.
When the vessel is depressurized, the operator initiates normal shutdown cooling by use of the RHR system.
Lf that system is unavailable because the valve on the KHR shutdown cooling suc tion line f ails to open, the operator initiates the alternate shutdown cooling mode.
For alternate shutdown cooling, the operator opens one S/RV and initiates eithat an RilR 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 steam lines and out of the S/RV and back to the suppression pool.
Cooling of the system is provided by use of an RHR heat axchanger.
As a result, an alternate cooling node is maintained.
In order to assure continuous long term heat removal, the S/RV 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 S/RV is required f or the alternate shutdown ooling mod'e, and no cycling of the S/RV was performed for the generic BWR S/RV operability test prodram.
The ability of the Brunswick S/RV to be extansively cycled for steam discharge conditions has been contirmed during steam discharge qualification testing of the valve by the valve vendor.
Based on the qualification teating of the S/RV's, the cycling of the valves in a controlled depressurinatica mode for steati discharge conditions will not
.idveraely af f ec t valve performance, and the prooability or the ytive ta iall open or ctosed is estremeif low.
d NRC QUESTION NO. 6:
l Describe how the values of valve C 's in report NED6-24988-P will be used at y
Brunswick.
Show that the methodology used in the test progras ;o determine the valve C will be consistent with the application at Brunswick.
y CP&L Response:
The flow coef ficient, C, for the Target Rock two-stage 7
safety / relief valve (S/ V) utilized in Brunswick was determined in the generic S/RV test program (NEDE-24988-P).
The average flow coefficient calculated f rom the test results for the Target Rock two-stage S/RV is reported in Table 5.2-1 of NEDE-24988-P.
This test value has been used by Carolina Power & Light to confirm that the liquid discharge flow capacity of the Brunswick S/RV's will be sufficient to remove core decay heat when injecting into the reactor pressure vessel (RPV) in the niternate shutdowa cooling mode.
The Cy value determined in the S/RV test demonstrates that the I
Brunswick S/RV's are capable of returning the flow injected by the RHR or CS pump to the suppression pool.
If it was necessary for the operator to place the Brunswick plant in the alternate shutdown cooling mode, he would assure that adequate core cooling was being provided by monitoring the following parameters:
RHR or CS flow rate, reactor vessel pressure, and reactor vessel temperature.
The flow coef ficient for the Target Rock two-stage valve reported in NEDE-24988-P was determined f rom the S/RV flow rate when the valve inlet was pressurized to approximately 250 psig.
The valve flow rate uas measured with the supply l
I line flow venturi upstream of the steam chest. The C for y
the valve was calculated using the nominal measured pressure differential between the valve inlet (steam chest) and 3 feet downstream of the valve and the corresponding measured flow I
rate.
Furthermore, the test conditions and test configuration were representative of Brunswick plant conditions for the alternate shutdown cooling mode, e.g.
pressure upstream of the valve, fluid temperature, f riction losses and liquid flowrate.
Therefore, the reported Cy values are appropriate for application to the Brunswick Plant.
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4
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