ML20072B571

Forwards Addl Info Re safety-relief Valve Operability Test Required by NUREG-0737,Item II.D.1, Reliability Engineering, Per NRC 830105 Request

ML20072B571
Person / Time
Site:	Peach Bottom
Issue date:	03/03/1983
From:	Daltroff S PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	Stolz J Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM NUDOCS 8303070071
Download: ML20072B571 (19)

Text

e PHILADELPHIA ELECTRIC COMPANY 23O1 M ARKET STREET P.O. BOX 8699 PHILADELPHI A. PA.19101 SHIE LDS L. DALTROFF stscraicraoo csom March 3, 1983 Docket Nos. 50-277 50-278 r

Mr. J. F. Stolz, Chief Operating Reactors Branch #4 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555

SUBJECT:

NUREG-0737, Item II.D.1, Safety-Relief Valve Operability Test

Dear Mr. Stolz:

The attachment to this letter provides the information requested in your letter of January 5, 1993 (J. F. Stolz to E. G.

Bauer, Jr., PECo.) regarding the Safety-Relief Valve Operability

Test required by NUREG-0737, Item II.D.l.

Should you have any questions regarding this matter, please do not hesitate to contact us.

Sincerely, A

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, ' Sill ~l } [

Attachment cc: Site Inspector Peach Bottom 8303070071 830303 PDR ADOCK 05000277 P PDR ON

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Docket Nos.- 50-277 50-278 s Attachment Peach Botton Atomic Power Station NURRG-0737, Item II.D.1 NRC OUESTION 1 l

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The test program utilized a " rams head" discharge pipe configuration. Peach Bottom utilizes a " tee" quencher configuration at the end of the discharge line. Describe the

' discharge pipe configuration used at Peach Bottom and compare the anticipated loads on valve internals in the Peach Bottom configuration to the measured loads in the test program. Discuss the impact of any differences in loads on valve operability.

1 RESPONSE TO OUESTION 1 i

The safety / relief valve discharge piping configuration at Peach Bottom utilizes a " tee" quencher at the discharge pipe exit. The average length of the eleven SRV discharge lines (SRV DL) is 129' and the submergence length in the suppression pool is approximately 8.58 The SRV test program utilized a ramshead at the discharge pipe exit, a pipe length of 1128 and a submergence length of approximately 138 Loads on valve internals during the test program are larger than loads on valve l internals in the Pe&ch Bottom configuration for the following reasons: '

1. No dynamic mechanical load originating a+ the " tee" quencher is transmitted to the valve in the Peach Bottom configuration because there is at least one anchor point

between the valve and the tee quencher.

2. The first segment length of the SRV piping that would be utilized for the alternate shutdown cooling mode at Peach Bottom averages less than 4 feet whereas the test facility first segment is 12 feet long thereby resulting l in a bounding dynamic mechanical load on the valve in i

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. e the test progras due to the larger moment arm between the SRV and the first elbow. 3

3. Dynamic hydraulic loads (backpressure) are experienced i by the valve internals in the Peach Bottom configuration. The backpressuro loads may be either (i) transient backpressures occurring during valve ,

actuation, or (ii) steady-state backpressures occurring l 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 mariaized by utilizing a submergence of 13', which is greater than Peach Bottom and a pipe length of 112' which is less than Peach Bottom.

The maximum transient backpressure occurs with high pressure steam flow conditions. The transient backpressure for the alternate shutdown cooling mode of operation is always much less than the design for steam flow conditions because of the lower upstream pressure and the longer valve opening time.

(b) The steady-state backpressure in the test program was maximized by utilizing an orifice plate in the SRVDL above the water level and before the ramshead. The orifice was sized to produce a backpressure greater than that calculated for any of the Peach Bottom SRVDL's.

An additional consideration in the selection of the ramshead for the test facility was to allow more direct muasurement 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.

The differences in the line configuration between the Peach Bottom plant and the test program as discussed above result in loads on the valve internals for the test facility which bound the actual Peach Bottom loads.

. l NRC OUESTION 2 f s

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 Peach Botton and compare the anticipated loads on valve internals for the Peach Bottom pipe supports to the measured loads in the test program. Describe the.

impact of any differences in loads on valve operability.

RESPONSB TO QUESTION 2 The Peach Bottom safety-relief valve discharge lines (SRV DL 's) are supported by a combination of snubbers, rigid supports, and spring hangers. These 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 locations of snubbers and rigid supports at Peach Botton are such that the loce tion of such supports in the BWR generic test facility is prototypical, i.e., in each case (Peach Bottom and the test f acility) there are supports near each change of direction in the pipe routing. Each SRYDL at Peach Bottom has 3 or 4 spring hangers, all of which are located in the drywell.

The dynamic load effects on the piping and supports of the test facility due t'o the water discharge events (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 features pertinent to this issue. Furthermore, analysis of a typical Peach Bottom SRVDL configuration has contirmed the applicability of the generic statement to Peach Bottom.

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 i ,

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these supports due to the weight of the water during the alternate shutdown cooling mode of operation. Therefore, design s 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 saubbers and rigid succorts, the dynamic loads resulting f rom liquid discharge during the alternate shutdown cooling mode of operation are significantly lower than those from the high pressure steam discharge. The spring hangers have been reviewed for the deflections resulting from the steam discharge dynamic event and found to be acceptable. In addition, the spring hangers have been evaluated for the increased dead load due to a water filled condition.

Both the spring hangers and piping stresses were acceptable.

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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l NRC QUESTION 3

.s Report NEDE-24938-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 encounted during the program.

RESPONSE TO OUESTION 3 No functional deficiencies or anomalies of the safety relief or relief valves were experienced during the testing at Wyle Laboratories for compliance with the alternate shutdown cooling mode requirement. All of the valves opened and closed without loss of pressure integrity or damage during all test runs, whether the run was valid or invalid. Anomalies encountered during the test program were all due to failures of test tacility instrumentation, equipment, data acquisition equipment, or deviation from the approved test procedure.

The test specification for each valve required six runs.

Under the test procedure, any anomaly caused the test run to be judged invalid. All anomalies were reported in the test report.

The Wyle Laboratories test log sheet for the Target Rock 3-Stage valve tests is attached. This valve is used in the Peach Bottom Atomic Power Statioh.

Each Wyle test report for the respective valve 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 data presented in Table 4.2-1 of NEDE-24986-P for each valve were obtained from Table 2.2-1 test runs and were based upon the selection criteria of:

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(a) Presenting the maximum respresentative loading I information obtained from the steam run data, - l (b) Presenting the maximum respresentative water loading information obtained from the 150F subcooled water test data, (c) Presenting the data on the only test run performed for the 500F subcooled water test condition.

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PAOE NO. 9 kh j TEST REPORT NO. 17476-03 s g Revision A N

TABLE I II TEST LOG FOR SRV TR-2 TEST TEST LOAD LINE TEST

. N0. MEDIA C0htlGURATl0N DATE REMARKS

201 Steam I 3/10/81 Back pressure low. Test Unacceptable.

. 202 Steam I 3/10/81 Installed 6.8" orifice.

Test Acceptable.

203 Water 1 3/10/81 Test Acceptable.

.' 204 Steam I 3/11/81 Test Acceptable.

, 205 Water i 3/11/81 Pipe loads high. See

, N0A # 5.

, 206 Steam I 3/11/81 Test Acceptable.

207 Water i 3/11/81 Not Acceptable. Low steam chest pressure.

. 208 Water i 3/11/81 Test Acceptable. Water

, temperature low.

3 209 Water i 3/30/81 Test Acceptable.

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210 Water i 3/30/81 Test Acceptable.

211 Water 1 3/30/81 Test Acceptable.

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i WYLE LABORATORIES l Huntsyd6e F4C334fy

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- PAGE NQ. [8 h TEST REPORT NO. 17476-03 '

Revision A NOTICE OF ANOMALY NOTICE NO. 5 P. O. MUMBER: 205-XH212 WYLE JOS NO. 17476-01 c CONTR ACT NUMBER: N/A CATEGORY: O SPECIMEN O PROC 9 DURE D TEST EQUIPMENT DATE: 3/u/A1 TO: General Electric CornpanY ATTN: Mr. R. Miller c PART NAME. Target Rock 3-Stage SRV _PART NO. N/A ,,

TEST: Low Pressure Water 1. D. NO. TR-2 SPECIFICATION: WTP 17450-01 PARA.NO. N/A NOTIFICATION MADE TO: J. Mross/A. Sallman DATE: 1/14/81 NOTIFICATION MADE BY: t- Mill m < VIA: Verbal e

REQUIREMENTS:

N/A c e

DESCRIPTION OF ANOMALY: e When the water control valve was opened to initiate the test, the entire system was subjected to a shock wave similar to water hammer. As a re.sult, loads of approximately c 10,000 and 16,000 pounds were observed at Struts I and 2. Review of the recorded data showed no abnormal pressure in the discharge line, but did show sharply varying pres-sure in the steam chest' and inlet water pipe.

DISPOSITION - COMMENTS - RECOMMENDATIONS:

The recorded data shows that the anomaly occurred in the inlet piping and/or stcom chest and, therefore, was not caused by the SRV. The probable cause was the forming of vapor "

in the inlet pipe because of the higher water temperature (233'F) and the low pressure (8 to 10 psig). The vapor then compressed when subjected to the higher pressure water (300 psigJ, thus causing a shock wave in the water system. Since the discharge pipe x loads were caused by the shock wave rather than the SRV, the data must be considered invalid.

5 The test was not repeated. However, three other water tests were conducted on this SRV, ,

and all data was consistent, in addition, water tests were performed on a two-stage Target Rock SRV, and no anomalies occurred, it ic, therefore, recommended that the test not be repeated.

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NRC OUES? ION 4 i

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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 Peach Botton for which the valves are required to operate and compare these plant conditions to the conditions in the test program.

Describc the plant features assumed in the event evaluations used to scope the test program and compare them to plant features at Peach Botton. For example, describe high level trips to prevent water from entering the steam lines ander high pressure operating conditions as assumed in the test event and compare them to trips.

used at Peach Bottom.

RESPONSE TO NRC OUESTION 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 referanced 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 conservatisa of a single active component failure or operator error postulated in the event sequence. It was concluded from this evaluation that the alternate shutdown cooling mode is the only expected event which will result in liquid at the valve inlet. Consequently, this was the event

! simulated in the S/RV test program. This conclusion and the test

[ results applicable to Peach Bottom are discussed below. The

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alternate shutdown cooling mode of operation will ha described in the response to NRC Question 5. ,

s The BWR Owners Group identified 13 events by evaluating the initiating events described in Regulatory Guide 1.70, aevision 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 13 events, only nine are applicable to the Peach Bottom plant because of its design and specific plant configuration. Four events, namely 5, 6, 10 and 13 are not applicable to the Peach Bottom plant for the reasons listed below:

(a) Events 5 and 10 are not applicable because Peach Bottom does not have a High Pressure Core Spray system.

(b) Event 6 is not applicable because Peach Bottom does not have'RCIC head sprays.

(c) Event 13 is not applicable because large breaks will not be isolated at Peach Bottom.

For the nine remaining events, the Peach Bottom 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 Peach Bottom because the base case ant'.ysis does not include any plant features which are notsaiready present in the Peach Bottom t design. For these events, Table 1 demonstrates that the Peach l Bottom specific features are included in the base case analysis presented in the BWR Owners Group submittal of September 17, i 1980. It is seen from Table 1, that all plant features assumed L in the event evaluation are also existing features in the Peach j Bottom plant. All features included in this base case analysis j are similar to plant features in the Peach Bottom design.

Furthermore, the time available for operator action is expected i to be longer in the Peach Bottom plant than in the base case l analysis for each case where operator action is required due to the conservative nature of the base case analysis.

Event 7, the alternate shutdown cooling mode of operation, is the only expected event which will result in liquid or two-phase i fluid at the S/RV inlet. Consequently, this event was simulated i in the BWR S/RV test prog ra m. In Peach Bottom, this event  !

involves flow of subcooled water (approximately 35 oF subcooled)

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at a pressure of approximately 75 psig. The S/RV inlet fluid conditions tested in the BWR Owners Group S/RV test program, as s documented in NEDE-24988-P, are 150 to 500 subcooled liquid at 20 psig to 250 psig. These fluid conditions envelope the conditions expected to occur at Peach Botton in the alternate shutdown cooling mode of operation.

As discussed above, the DWR Owners Group evaluated transients including single active f ailures that would marinize 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 Peach Bottom plant-specific fluid conditions expected for the alternate shutdown cooling mode of operation.

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NRC OUESTION 5 s

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 progras? 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 OUESTION 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 Peach Bottoa. 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 valres and removing heat through the main condenset. If the main condenser is unavailable, the operator could depressurize the reactor vessel by using the SRV's to discharge steam to the suppression pool. If SRV operation is required, the operator cycles the valves in order to assure that the cooldown rate is maintained within the technical specification limit of 1000F per hour. When the vessel is depressurized, the operator initiates normal shutdown coo' ling by use of the RHR system. If that system i is unavailable because the valve on the RHR shutdown cooling i

suction line fails to open, the operator initiates the alternate j shutdown cooling mode.

For alternate shutdown cooling, the operator opens an 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 steam linos and out of the SRV and back to the suppression pool.

Cooling of the system is provided by use of an RHR heat exch a nger. As a result, an alternate cooling mode is maintained.

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In order to assure continuous long term heat removal, the SRV is kept open and no cycling of the valve is performed. In order ,

to control the reactor vessel cooldown rate, the operator is instructed to control the flow rate into the vessel.

Consequently, no cycling of the SRV is required for the alternate shutdown cooling mode, and no cycling of the SRV was performed

'for the generic BWR SRV operatility test program.

The ability of the Peach Dottom SRV to be extensively cycled for steam discharge conditions has been confirmed during steam ,

discharge qualification testing of the valve by the valvo vendor.

Based on the qualification testing of the SRV8s, the cycling of the valves in a controlled depressurization mode for steam discharge conditons will not adversely affect valve performance and the probability of the valve to fail open or closed is extremely low.

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.NRC OUESTION 6 s 3 Describe how the values of valve Cv's in report NEDE-24988-P will be used at Peach Bottoa. Show that the methodology used in the test program to determine the valve Cy will be consistent with the application at Peach Bottom.

RESPONSE TO NRC OUESTION 6 The flow coefficient, Cv, for the Target Rock 3-Stage safety relief valve (SRV) utilized in Peach Botton was determined in the generic SRV test program (NEDE-24988-P) . The average flow coefficient calculated from the test results for the Target Rock Valve is reported in Table 5.2-1 of NEDE-24988-P. This test value has been used by Philadelphia Electric Company to confirm that the liquid discharge flow capacity of the Peach Bottom SRV's will be sufficient to remove core decay heat when injecting into the reactor pressure vessel (RPV) in the alternate shutdown cooling mode. The Cy value determined in the SRV test demonstrates that the Peach Botton SRY's are capable of returning sufficient flow to the suppression pool to accomodate injection 1

by the RHR or CS pump.

If it were nece'ssary for the operator to place the Peach Botton plant in the alternate shutdown cooling mode, he would assure that adeguate core cooling was being provided by monitoring the following parameters: RHR or CS Flow rate, reactor 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 CV for the valve was calculated using the l nominal measured pressure ditferential between the valve inlet l (steam chest) and 3' downstream of the valve and the l corresponding measured flowrate. Furthermore, the test

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, l conditions and test configuration were. representative of Peach Bottom plant conditions for the alternate shutdown cooling mode, s e.g. pressure upstream of the valve, fluid temperature, f riction losses and liquid flowcate. Therefore, the reported Cy values ,

are appropriate for application to the Peach Bottom plant. {

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