ML20069H802
| ML20069H802 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 04/04/1983 |
| From: | Causey J GEORGIA POWER CO. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM IEB-79-14, NED-83-237, TAC-44586, TAC-44587, NUDOCS 8304120225 | |
| Download: ML20069H802 (41) | |
Text
P Georg a Power Compan/
333 Ptedmont Avenue Atlanta, Georgia 30308 Telephone 404 5261895 Maing Address:
Post Office Box 4545 Atlanta, Georg:a 30302 h
Georgia Power J. C. Causey V.ce President Fossa and Hydro Generat on NED-83-237 April 4, 1983 Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555 NRC DOCKETS 50-321, 50-366 CPERATItG LICENSES DPR-57, NPF-5 EDWIN I. HATCH NUCLEAR PLANT UNITS 1, 2 NUREG-0737 ITEM II.D.1 SAFETY RELIEF VALVE (SRV) TESTING Gentlemen:
Your letter dated December 29, 1982 requested Georgia Powcr Company (GPC) to provide additional information related ~ to the SRV test program.
The attached report provides responses to the six questions from your letter pertaining to the subject item.
Note that question 2 of the response requires a dead weight analysis of the Hatch SRV discharge lines to confirm that results of the test program cpply to Plant Hatch. This analysis is being perfurmed for GPC by Bechtel Power Corporation to meet the requirements of I&E Bulletin 79-14 and the Mark I Long-Term Plan. It is anticipated that we will receive a report of this analysis by July 29, 1983, and upon receipt, the report will be submitted to NRC as a supplementary response.
If you require additic:ial information regarding this response, please contact this office.
Very truly yours, L[
S 8304120225 830404' / / yQ 6'C. Causey//fg6 PDR ADDCK 05000321 P PDR CT/mb Enclosure i
xc: J. T. Beckham, Jr.
H. C. Nix, Jr.
J. P. O'Reilly (NRC- Region II)
. Senior Resident Inspector L.
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E. I. HATCH NUCLEAR PLANT UNIT I RESPONSES TO NRC QUESTIONS RELATIVE TO SRV TESTING 0
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,NRC QUESTION 1 The test program utilized a " rams head" discharge pipe configuration.
E. I. Hatch utilizes a " tee" quencher configuration at the end of the discharge line. Describe the discharge pipe configuration used at E. I. Hatch and compare the anticipated loads on valve internals in the
! E. I. Hatch configuration to the measured loads in the test program.
Discuss the impact of any differences in loads-on valve operability.
1 RESPONSE TO QUESTION 1 The safety / relief valve discharge piping configuration at E. I. Hatch utilizes a " tee" quencher at the discherge pipe exit. The average length' of the 11 SRV discharge lines (SRVDL) is 107'-9 11/16" and the submergence
, length in the suppression pool is approximately - 7'-8". The SRV test program 3 utilized a ramshead at the discharge pipe exit, a pipe length of 112' and a
.. submergence length of approximately 13'. Loads on valve internals during the test program are larger than loads on valve internals lbs the E. I. Hatch configuration for the following reasons:
I 1. No dynamic mechanical load originating at the " tee" quencher is transmitted to the valve in the E. I. Hatch configuration because
? there is at least one anchor point between the valve and the tee quencher.
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4 2. The first length of the segment of piping downstream of the SRV in
[ the test facility was longer than .the E. I. ' Hatch piping, thereby resulting in a bounding dynamic mechanical load on the valve in the
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test program due to the larger moment arm between the SRV and the 1
first elbow.- The first segment length in the-test facility is 12 ft.
whereas this length in the E. I. Hatch configuration is given below..
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- Vent Length Vent Length A.
12' - 11 3/8"* F 2' - 5 5/8" ,
B 3' - 4k"- G 7' =4 5/8" C 7' - 10 5/8" H 12' - 11 5/8"* ,
D l' -'S 5/8"- 'J 2' - 8" E l' - 3" K 2' - 10 5/8" L 7' - 9"
- The first segment length in the test facility does not include the elbow length. The above first segment length for 'the vents include the elbow
' .j length. Therefore by subtracting the elbow length from vents A and H all the first segment lengths would be less than 12 feet.
- 3. Dynamic hydraulic loads (backpressure) are experienced by the valve internals in the E. I. Hatch configuration. 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 i upstream pressure, short'er valve opening times,. greater line.
submerge 1ce, and smaller SRVDL air volume. .The transient backpressure in the test program was maximized by utilizing a submergence of 13', and a pipe length of 112'. 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.
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1 (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 t .
produce a backpressure greater than that calculated for any-of the'E. I. Hatch SRVDL's.
The differences in the line configuration between the E. I. Hatch plant and the test program as discussed above result in the loads on the. valve internals.for the test facility which bound the actual E. I. Hatch loads.
a . An additional consideration in the selection of the ramshead for the test l 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 f 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 the E. I. Hatch
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and the test facility will not have any adverse effect on SRV operability
, at E. I. Hatch relative to the test facility.
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NRC QUESTION 2 The test configuration utilired no spring hangers as pipe supports.
Plant specific configurations do use spring hangers in conjunction with snubber and' rigid supports. Describe the safety relief valve pipe
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supports used at E. I. Hatch and co= pare the anticipated loads on valve
"} internals for the E. I. Hatch pipe supports to the measured loads in the 1
. .j-test program. Describe the impact of any differences in loads on valve operability. -
3-RESPONSE T0"QUESTIdN 2
, The E. I. Hatch safety-relief valve discharge lines (SRVDL's) are supported by a co=bination of snubbers, rigid supports, and spring hangers. The locations of snubbers and rigid supports at E. I. Hatch are such that the location of such supports in the BWR generic test facility is prototypical.-
i.e. , in each case (E. I. Hatch and the test facility) there are . supports near each change of direction in the pipe routing. Additionally, each SRVDL at E. I. Hatch has only 1 to 2 spring hangers, all of which are located in the drywell. The spring hangers, snubbers, and rigid supports were designed to accc==odate co=binations of loads resulting from piping dead weight, thermal canditions, seismic and suppression pool hydrodynamic events, and a high pressure steam discharge transient.
The dyna =ic 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 EWR's since the test facility was designed to be prototypical of the features pertinent to this issue.
Further=cre, analysis of the E. I. Hatch SRVDL configurations will be-perforned to confirm the applicability of this conclusion to E.1. Hatch, c
During the water discharge transient there will be significantly' lower dynamic loads acting en the snubbers-and' rigid supports than during the
,I stes= discharge transient. This will-= ore than offset.the small increase 4
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l 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 1s assured as they are
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j' designed for the larger steam discharge. transient loads.
l This question addresses the design adequacy of the spring ha:- vith 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 d lower than those from the high pressure steam discharge. Therefore, it is believed that sufficient margin exists in the E. I. Hatch 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 SRVs to open to establish the alternate shutdown cooling path since the loads occur in the SRVDL only after valve opening.
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4 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.
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4 RESPONSE TO QUESTION 3 IL lF 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 instru-mentation, 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 Target Rock Model 7567F valve tests is attached. This valve is used in the E. I. Hatch Nuclear Power Station.
4 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 l} and states the reason'for considering the run invalid. No anomaly
! encountered during the required test program affects any valve safety j or operability function.
, All valid test runs are identified in Table 2.2-1 of NEDE-24988-P. The i- data presented in Table 4.2-1 for each valve were~obtained from the j
i Table 2.2-1 test runs and were based upon the selection criteria of:
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- , (a) Presenting the maximum representative loading information obtained t
from the steam run data, I
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- (b). Presenting the maximum representatiOe water' loading information-obtained-from the 15'F subcooled water test data',
(c) Presenting the data on the only test run performed for the 50*F subcooled water test condition.
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1 TE51 REPORI 17476-04 l'
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'REPcRT 30_
1 WYLE LABOMTUMS ,
ouR ;Ca NO.
17476 i sCIENitFIC SERvlCES ANo' SYSTEMS GROUP ,
HUNTSVILLE, ALAB AMA ~ YOUR P. O. NO. 205-XH212 CONTRACT N/A General Elechr. ic . Company
. 175 Curtner Aver,ue .
PAGE 1 of 77 PAGE REPC
. San Jose, California
.- b .d DATE MAY 8. 1981 Januarv 18, 1 Revision A:
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1.0 PURPOSE '-
i The purpose of this report is to present the req'uirements, procedures, and results of steam and low pressure .wate'r operability tests performed on a j
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Target Rock 6X10 Safety Relief Valve (SRV) identified as TR-1. The tests I were performed to determine if the SRV would operate properly wh'en sub- l jected to the test conditions specified in General Electric Specification 22A7424, Revision B. ,
4
2.0 REFERENCES
2.1 General Electric Purchase Order 205-XH212.
2.2 Vyle Laboratories' Test Procedure 'No. 17450-15 2.3 Vy.le Laboratories' Test Procedure No. 17450-01.
2.4 General Electric Specification 22A7424, Revision B.
2.5 Vyle Laboratories' Report No. 17450-02. ,
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2.6 Vyle Laboratories' Test Report No. 45503-04. .
i 2.7 Vyle Laboratories Test Report No. 45503-07.' -
30 MANUFACTURER t
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1 Target Rock Corporation 1966E Broadhollow Road East Farmingdale, NY ,
11735 .
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gep035 Jnc $3yS f tte %!Of" **On Conta fsec .n that report es the result Of Complete trtd c.iettwily CGndWCit$ O el to in D cf his know6eoge true and correct an ,f fp i
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TEST LCG FOR SRV TR-1 -
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fa ,. 301 steam I 3/17/81 Acceptable ,_
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3/17/81 GN Regulator failed. I' l- Da!anotacceptable. }
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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 E. I. Hatch 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 evalu'ations used to scope' the test program and compare them to plant 2 - features at E. I. Hatch. 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.
E. I. Hatch.
RESPONSE TO NRC QUESTION 4 The purpose of the S/RV test program was to demonstrate that the Safety .j 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 conven-tional 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; expected event which will result in liquid at the valve inlet. Conse-quently, this was the event simulated in the S/RV test program. This
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conclusion and the test results applicable to E. I. Hatch are discussed below. The alternate shutdown cooling mode of operation has been described in the response to NRC Question 5.
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The S/RV inlet fluid conditions. tested in the BWR Owners Group S/RV test.
l program, as documented.in N p E-24988-P, are 15* to 50* subcooled liquid-at 20 psig to-250 psig. These fluid conditions' envelope the conditions
-expected to occur at E. I.' Hatch in thefalternate shutdown cooling mode of operation..
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A 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 mitigate these events, are summarized in Table"1. Of these j 13 events, only 10 are applicable to the E. I. Hatch plant because_of f- its design and specific plant configuration. 3 events, namely 5, 6, & 10 are-not applicable to the E. I. Hatch. plant for the reasons-listed below:
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(a) Events 5 and 10 are not applicable, because Plant E. I. Hatch does not have a HPCS system.
4 (b) Event 6 is not applicable because Plant E.' I. Hatch does not
- have RCIC head sprays.
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. For the 10 remaining events, the E. I. Hatch specific features, such' as trip logic, power supplies, instrument line configuration, alarms and i , 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 E. I. Hatch because the' base case analysis does not include any plant.
] features which are not already present.in the E. I. Hatch design. For 3
these events, Table 1 demonstrates that the E. I. Hatch specific features: ,
- are included in the base' case analyses presented dxt the BWR Owners Group submittal of September 17, 1980. It is seen~ from Table 1, that all plant features assumed-in the event evaluation arefalso existing features.in the E. I. Hatch plant. All-features included in this. base case analysis
- l. are similar to plant features in the'E. I. Hatch design. Furthermore, the
- -time.available-for operator action is expected to be longer in the
, lE. I. Hatch plant ~than in the base case analysis for each case where-operator action,is required.
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- 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 E..I. Hatch, this event-involves flow of subcooled water.
.(approximately 130*F subcooled) at a pressure of approximately 85'psig.
-The test conditions clearly envelope these plant con ditions.
As discussed above, the BWR Owners Group evaluated transients including b 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 E. I. Hatch 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 depres-surization mode in a plant-specific application. Was this mode simulated l 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?
l RESPONSE TO NRC QUESTION 5 4
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 E. I. Hatch. 'The sequence of events l leading to the alternate shutdown cooling mode is given below.
r 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 l
i operator could depressurize the reactor vesse1~by using the SRV's to l discharge steam to 'the suppression pool. If SRV operation is required, l
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. 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 RER 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 SRV and back to the suppression pool. Cooling of the system is provided by use of an RER 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-a t
i g
4 . . .- .-
L 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. l l - .The ability of the E. I. Hatch SRV to be extensively. cycled for steam
- discharge conditions has been confirmed during steam discharge qualifi-cation testing of the valve by the valve vendor. Based on the qualifi-cation testing of the SRV's, the cycling of the valves in a. controlled ;
depressurization mode for steam discharge conditions will not adversely ;
I
- x. affect' valve performance and the probability of the valve to fail open
- or closed is extremely low, i
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NRC QUESTION 6 Describe how the values of valve C 's v in report _NEDE-24986-? Will be used a: E. I. Eatch. Show that the =ethodology used in the test program to deter =ine the valve C , will be censistent with the application at E. I Hatch.
RESPONSE TO NRC QUESTION 6 J 3 The flow coefficient. C , for.the Target Rock Model 7567F safety relief V
valve (SRV) utilized in E. I. Eatch was datermined in. the generic SRV test progra= (NEDE-24988-?). The average flow coefficient calculated frc= the test results for the Target Rock Model 7567F, is reported in Table 5.2-1 of NEDE-24986-P. This test value has been used by Georgia
?cuer t'o confir= that 'the liquid discharge flow capacity of the E. I.
Eatch SRV's will be sufficient to remove core decay heat when injecting into the reactor pressure vessel (R?V) in the alternate shutdown cooling I
= ode. The Cv value deter =ined in the SRV test de=enstrates that the E. I. Eatch SRV's are capable of returning the flow injected by the RER or CS pu=p to the suppressio'n pool.
b
. If it were necessary for the operater to place the E. I. Eatch plant in ',
~ the alternate shutdown cooling mode, he would assure that adequate core cooling was being provided by monitoring the following para =eters: RER or CS flow rate, reactor vessel-pressure and reactor vessel te=perature.
The flev coefficient fer the Target Rock Model 7567F valve reported in NIDE-24988-? was deter-ined frc= the SRV flow rate when the valve inlet was pressurized to approni=ntely 250 psig. The valve flow rate was =ea-sured with the supply line flev venturi upstrea= of the steam chest.
The C , f cr the valve was calculated using the nc=inal =casured pressure dif ferential between the valve inlet (steam chest) and 3' dcunstren= of the valve. cad the ccrrescading =casured flowrate. Furthermore, the test cendi icas and test cenfiguration were representative of E. I. Estch plant cenditiens for the alternate shutdeva cooling = ode, e.g. pressure upstren=
of the valve, fluid te=perature, friction losses and liquid fleurate.
Therefcre, the reported C 9 values are apprcpriate for applicatien'to the
{I E. I. Earch plant.
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REFERENCES.
- l. F. L. Leverenz, "Probabilistic Evaluation of High Pressure Liquid-Challenge of Safety / Relief Valve-Piping " Science Applications Inc.
Palo Alto, California, April 1981.
. November 14, 1979.
- l. September 17, 1980.
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E. I. HATCH NUCLEAR PLANT UNIT 2 RESPONSES TO NRC QUESTIONS RELATIVE TO SRV TESTING n
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NRC QUESTION 1 The test program utilized a "ramshead" discharge pipe configuration.
E. I. Hatch utilizes a " tee" quencher configuration at the end of the discharge line. Describe the discharge pipe configuration used at E. I. Hatch and compare the anticipated loads on valve internals in the E. I. Hatch configuration to the measured loads in the test program.
Discuss the impact of any differences in loads on' valve operability.
i .
RESPONSE TO QUESTION 1 The safety / relief valve discharge piping configuration at E. I. Hatch utilizes a " tee" quencher at the discharge pipe exit. The average length of the 11 SRV discharge lines (SRVDL) is 109'-2 7/8" and the submergence length in.the suppression pool is approximately 7'-8". The SRV test program utilized a ramshead at the discharge pipe exit, a pipe length of 112' and a submergence length of approximately 13'. Loads on valve internals during the test program are larger than loads on valve internals in the E. I Hatch configuration for the following reasons:
- 1. No dynamic mechanical load originating at the'" tee" quencher is transmitted to the valve in the E. I. Hatch 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 E. I. Hatch 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 ft.
. whereas this length in the plant configuration is given below:
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n Vent "A" - 19 5/8" Vent "E" - 19 5/8" Vent "K" - 4'O
~
Vent B -
2'-3" Vent F -
4' Vent. L - 22k" Vent C -
~ 4'-0 Vent G -
4'-0. . Vent M - 22k" Vent D.- 5'-10" Vent H - 19 5/8" l' 3. Dynamic hydraulic loads (backpressure) are experienced by the valve
~
internals in the E. I. Hatch configuration. The backpressure loads may be either (i) transient backpressures occurring during valve -
L
!l. actuation, ~ or (ii) steady-state backpressures occurring during j 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 l
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 maximiz'ed by utilizing a submergence of 13', and a pipe length of 112'. The maximum transient backpressure occurs with high pressure steam flow conditions. The transient backpressure for the alternate shutdown cooling mode of operation $s always much less than the design for steam flow conditions because of the lower up-stream 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 I the SRVDL's.
The differences in the line configuration between the E. I. Hatch plant and the test program as discussed above result in the loads on the valve internals for the test facility which bound the actual E. I. Hatch loads.
An additicnal consideratio'n 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
' l :. . . . .
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^
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 E. I. Hatch and the
- test facility will not have any adverse.effect on SRV operability at E. I. Hatch relative to the test facility.
- P.
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NRC OUESTION 2 The test configuration utilized no spring hangers as pipe supports.
Plant specific configurations do use spring-hangers in conjunction with snubber and rigid supports. Describe the safety relief valve pipe supports used at E. I. Hatch and compare the anticipated loads on valve .
~'
internals for the E. I. Hatch pipe supports to the measured loads in the q test program. Describe the impact of any differences in loads on valve
~ ' ~
operability. -
RESPONSE T0" QUESTION 2 The E. I. Hatch safety-relief valve discharge lines (SKVDL's) are supported by a cc bination of snubbers, rigid supports, and spring hangers. The locations of snubbers and rigid supports at E. I. Hatch are such that the location of such supports in the BWR generic test facility is prototypical, i.e. , in each case (E. I. Hatch and the test facility) there are . supports near each change of direction in the pipe routing. Additionally, each SRVDL at E. I. Hatch has only 1 to 2 spring hangers, all of which are.
located in the drywell. The spring hangers, snubbers, and rigid supports were designed to acco=modate combinations of loads resulting from piping dead weight, thermal conditions, seismic and suppression pool hydrodynamic 4
events, and a high pressure steam discharge transient.
The dyne =ic 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 ERR's since the test facility was designed to be prototypical of'the features pertinent to this issue.
Further=cre, analysis of the E. I. Hatch SRVDL configurations will be performed to confirm the applicability of this conclusion to E. I. Hatch.
1 During the water discharge- transient there vill be significantly lower l dyna =ic loads acting en the snubbers and' rigid supports than during the-stea= discharge transient. -This vill more than offset the small ircrease j.!
j l :ik A. l g .#.3. .
I 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.
1.
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 E. I. Hatch 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 SRVs to open to establish the alternate shutdown cooling path since the loads occur in the SRVDL only after valve opening.
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4 I
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NRC QUESTION 3
. l 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 f
l encountered during the program, i
RESPONSE TO QUESTION 3 l
l l 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 instru-mentation, 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 Target Rock Model 7567F valve tests is attached. This l
I valve is used in the E. I. Hatch Nuclear Power 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 af fects any valve safety or operability function.
All valid test runs are identified in Table 2.2-1 of NEDE-24988-P. The l 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.
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(b)' Presenting the maximum representative water loading information obtained fran 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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JE51 REPORI 17476-04 i
/ ' REPORT NO._
k $3 NEbb '
OUR JCB NO.
17476
' SCIENTIFIC sERvlCES A'40 SYSTEMS GROUP . .
< HUNTsVILLE ALAB AMA
- YOUR P. O. NO. 205-XH212 CONTRACT _ N/A i
General Elec'tric Company 175 Curtner Avenue .
PAGE 1 of 77 PAGE REPO l San Jose, California
! .d DATE H5Y 8. 1981-
. Revision A: Januarv 18. 1 a.
I -
. 1.0 PURPOSE '
4
} The purpose of this report is to present the req'uirements, procedures, and results of steam and low pressure . water operability tests performed on a
~
Target Rock 6X10 Safety Relief Valva (SRV) identified as TR-1. The tests were performed to determine if the SRV would operate properly wh'en sub-l Jected to the test conditions specified in General Electric Specification j < 22A7424, Revision B. , ,
j .g
) ' 2!O . REFERENCES ,.
I- 2.1 General Electric Purchase Order 205-XH212. -
7
)a 2.2 Wyle Laboratories' Test Procedure'No. 17450-15 2.3 Vy.le Laboratories' Test Procedure No. 17450-01.
f' '
- i. 4 General Electric Specification 22A7424, Revision B.
l{ , ,
l, 2.5 Vyle Laboratories' Report' No. 17450-02.
i 2.6 Vyle Laboratories' Test Report No. 45503-04.
I , 2.7 Wyle Laboratories Test Report No. 45503-07. '
j .%
I 30 MANUFACTURER
) ,I -
j' Target Rock Corporation !
1966E Broadhollow Road East Farmingdale, NY , 11735
. 1
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STATE 6F ALA8AMA t """***9'"'""'**9"""*VN***
, ' '= '
COUNTY oF MAotSoN , I T, " Y STEAM SERVICES L a r rv E . FraH er , w,ng a y,, ,,,,,,,,
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.-. - - - - - - . ~
PAGENO. B . i . ,.
,i TEST REPCRT NO. 17476-04 .
- - - i t
TABLE I ;' s
- TEST LOG FOR SRV TR-1 .
5
. s .
J.- Test Test , Load Line Test' -
'& -I, ,
No. Media Configuration Date Remarks t
+
l e . 301 steam I '/17/81 3 Acceptable , j 302 Vater 1 3/I'7/81 GN., Regulator failed. , t' Data not acceptable. } i 303 vater 1 3/17/81 Acceptable , i f 304 Steam I 3/17/81 Acceptible I., i
- 305 water, . I 3/18/81 Acceptable <
i i
- 4 Steam I 3/18/81 Acceptable l .
306 , , ! , 307 Vater , 1 3/18/81 Acceptable 308. Water i 3/18/81 special test at elevated j . te=perature and low pres- ~ j sure requested by G.E. ; 1 g
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4 NRC QUESTION 4
~ l 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 E. I. Hatch 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 E. I. Hatch. For exa'ple, m describe high level trips to prevent water from entering the steam lines under high pressure operating 4
conditions as assumed in tha test event and compare them to trips used at , E. I. Hatch. 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 4 would be maximized. Test pressures were the highest predicted by conven-tional 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 expected event which will result in liquid'at the valve inlet. Conse-quently, this was the event simulated in the S/RV test program. This conclusion and the test results applicable to E. I. Hatch-are discussed below. The alternate shutdown cooling mode of operation has been described in the response-to NRC Question 5. 1
, i
- v. * *
'j
~~
.n 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 E. I. Hatch 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 i 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 10 are applicable to the E. I. Hatch plant because of its design and specific plant configuration. 3; events, namely 5, 6, & 10 are not applicable to the E. I. Hatch plant for the reasons listed be' low: (a) Events 5 and 10 are not applicable, because Plant E. I. Hatch does not have a HPCS system. (b) Event 6 is not applicable because Plant E. I. Hatch does not i have RCIC head sprays. 4 For the 10 remaining events, the E. I. Hatch specific features, such as trip logic, power supplies. instrument line configura. tion, 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 f
has demonstrated that in each case..the base case analysis is applicable to E. I. Hatch because the base case analysis does not include any plant features which are not already present in the E. I.~ Hatch design. 'For-these a*:ents, Table 1 demonstrates that the E. I. Hatch 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 E. I. Hatch plant. All features included in this base case analysis-are similar to plant features in the E. I. Hatch design. Furthermore, the time available -for operator action is expected to. be longer in the .
E. I.. Hatch plant than in the base case analysis for each case where .l operator action is required. I l 1: , -.t ;. ,
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 E. I. Hatch, this event involves flow of subcooled water (approximately 130*F subcooled)' at a pressure of approximately 85 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 E. I. Hatch plant-
. specific fluid conditions expected for the alternate shutdown cooling modi of operation.
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e l l NRC QUESTION-5 The valves are likely~to be extensively. cycled in a controlled depres- , surization mode in a plant-specific application. Was this mode simulated j .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 4
] .The BWR safety / relief valve (SRV) operability test program was designed 1
j to simulate the alternate shutdown cooling mode, which is the only I expected liquid discharge event for E. .I. Hatch. The sequence of events
. leading to the alternate shutdown' cooling mode is-given below.
i Following normal reactor shutdown, the reactor operator depressurizes the 1 reactor vessel by opening the turbine bypass valves and removing heat i through the main condenser. If the main condenser is unavailable, the i operator could depressurize the reactor vessel by using the SRV's to i discharge steam to the suppression pool. If SRV operation is required, j the operator cycles the valves in order to assure that the cooldown rate i 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 RER system. If that system is unavailable because j > -
the valve on the RER shutdown cooling. suction line fails to open, the l operator initiates the alternate shutdown cooling mode. l l i For alternate shutdown cooling, the operator opens one SRV and initiates either an RER or core spray pump utilizing the suppression pool as the suction source. The reactor vessel is filled such that water is. allowed I to flow into the main steam lines and out of the SRV and back to the-I suppression pool. Cooling of the system is provided by use of an RHR l~ . ' heat exchanger. As a result, an alternate cooling mode is maintained.. 1 In order to assure continuous long term' heat removal,-the SRV is kept t 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 j= g.
J .
flow rate into.the. vessel. Consequently, no cycling of the SRV is l 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 E. I.' Hatch SRV to be extensively cycled for steam ! discharge conditions has been confirmed during steam discharge qualifi-cation testing of the valve by the valv,e vendor. Based on the qualifi-cation testing of the SRV's, the cycling of the valves in a controlled l, depressurization mode for steam discharge conditions will not adversely 'A ~
- f. affect valve performance and the probability of the valve to fail open or closed is extremely low.
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, NRC QUESTION 6 Describe how the values' of valve C 's in report NEDE-24988-P will be used at E. I. Hatch. Show that the methodology used.in the test program to determine the valve Cy will be consistent with the application at E. I Hatch.
RESPONSE TO NRC QUESTION 6 1
, The flow coefficient, Cy , for the Target Rock Model 7567F safety relief valve (SRV) utilized in E. I. Hatch was determined in the generic SRV test program (NEDE-24988-P). The average flow coefficient calculated l
from the test results for the Target Rock Model 7567F, is reported in Table 5.2-1 of NEDE-24988-P. This test value has been used by Georgia Power to confirm that the liquid discharge flow capacity of the E. I. Hatch SRV's will be sufficient to remove core decay heat when injecting I 'into the reactor pressure vessel (RPV) in the alternate shutdown cooling l mode. The Cy value determined in the SRV test demonstrates that the f E. I. Hatch 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 the E. I. Hatch plant in the alternate shutdown cooling mode, he would assure that adequate core
, cooling was being provided by monitoring the following parameters: RHR l* or CS flow rate, reactor vessel pressure and reactor vessel' temperature. -
l l The flow coefficient for the Target Rock Model 7567F valv,e 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 mea-sured 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 corresonding measured flowrate. Furthermore, the test-conditions and test confih;uration were representative of E. I. Hatch plant conditions for the ' alternate shutdown cooling mode, e.g. ' pressure upstream
.of the valve, fluid temperature, friction losses and liquid flowrate.,
Therefore, the reported C y values are appropriate for application to the
- - E. I. Hatch plant, i ..
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' REFERENCES.-
- 1. F. L. Leverenz, "Probabilistic Evaluation of High Pressure Liquid Challenge of Safety / Relief Valve Piping". Science Applications,.Inc.
Palo Alto, California, April 1981.
- 2. Letter to D. G. Eisenhut (USNRC) from T. D. Keenan (BWR Owners Group),
November 14, 1979.
- 3. Letter to R. H. Vollmer (USNRC) from D. B. Waters (BWR Owners Group),
September 17, 1980. t . t 9 9* O 4 4 2
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