ML20040H109
| ML20040H109 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 02/11/1982 |
| From: | Deacon W BOSTON EDISON CO. |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| 82-51, NUDOCS 8202170196 | |
| Download: ML20040H109 (9) | |
Text
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a BOSTON EDISON COMPANY mon Bovtsvon Svesse BohTON. MASSACHUSCTTs 02199 Februa ry 11, 1982 g\\
'9, BECo. Ltr. =82-51 Mr. Domenic B. Vassallo, Chief f
Operating Reactors Branch f2 2,
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Division of Licensing C
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Office of fluclear Reactor Regulation 6U84' Y
U.S. fluclear Regulatory Commission 6
Washington, D. C.
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cense fio. DPR-35 ocket flo. 50-293 D
c3 Report on Target Rock Safety / Relief Valves
References:
(a) Licensee Event Report 81-062/01T-0, Boston Edison Letter #81-271 of fiovember 27, 1981 (b) Decerber 18, 1981 fleeting, summarized in a memo from Mr. K. T. Eccleston to Mr. T. A. Ippolito
Dear Sir:
Boston Edison herein submits responses to the concerns involving Target Rock Safety / Relief Valves (SRV) raised by the fiRC at the December 18, 1981 meeti ng,
and reiterated in Reference (b).
Section I:
Description of Valve Failure and Corrective Actions Backoround The safety relief valves (SRV's) originally used in the Pilgrim plant were Model 67F 3 Stage Target Rock valves.
These valves were very sensitive to leakage past the 1st Pilot stage.
Leakage of the order of 15 lbs. per hour cculd cause the valves to self activate and hold them open until the reactor had depressurized.
The new 2 stage valves are much less sensitive to leakage past the pilot, and, because they were considered a design improvement which would reduce down time, they were purchased in late 1979 and installed during the PNPS 1980 refueling outage.
The SRV's sit on short standpipes on the steam lines. The exhaust lines from the valves are connected solidly to the Torus by runs of about 100 feet of 12 inch pipe. The valves are nominally 6" inlet and 10" outlet; all have main seats of about 5" diameter and a rated capacity of about 800,000 LBS. per hour. The nominal set point of the valves is 1095 11 psig. The 2 stage topworks were installed on reworked bodies at the Pilgrim site after the bodies of the original 3 stage valves had been re-worked at Wyle.
Prior to installation, the 2 stage topworks had been calibrated on the Wyle steam test facility. The assembled valves were first operated at start-up in May 1980.
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CO2 TON EDIEDN COMPANY Mr. Domenic B. Vassallo, Chief February 11, 1982 Page 2 In addition to operating in a relievina mode when steam pressure exceeds the set point, the valves also can operate similarly in the ADS mode.
In the ADS mode, air or nitrogen gas pressure is ducted through a -3 way solenoid valve to a diaphragm operator, which moves the second stage pilot disk, vents a cavity above
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the primary stage piston, and lifts the primary disk.
At start-up the actuating medium was nitrogen. This nitrogen came from a cryogenic tank outside the building. On or about October 31, 1980, the actuating medium was changed to instrunent air regulated to 120 psig because of malfunctions of the nitrogen supply system.
Operational History On July 25, 1980, valve D did not respond to several attempts to manually actuate it. Valves B and C did respond to an opening signal.
No attempt was made to open Valve A.
The solenoid on valve D was replaced after its disassembly revealed that the valve plunger _and disk were smeared with Loctite. After cleaning the Loctite from these parts, the solenoid was reassembled and functioned normally when tested. The solenoid valve on Valve A was also found to be leaking slightly through the bleed port, and was replaced.
During start-up, at a Reactor pressure of 160 psig, the D valve failed to open when its switch was manually actuated. Two additional actuation attempts were made and the valve opened each time.
No apparent reason was determined for this failure to open on the first try, but control room personnel were unsure if an open indication was received, or if the switching sequence was completed. Valve A opened satisfactorily and no leakage was evident from any solenoid valve during start-up.
On August 1, the D valve again did not open on a manual signal. The Reactor was shut down and the valve examined. The valve top works was removed, and the various components tested and exainined. The solenoid valves were cycled 15 times, actuated by 120VDC and air at pressures ranging from 20 to 125 psig, without any -
sign of malfunction. Leakage in excess of.5 SCFH was noted through the solenoid valve, although some of this was through the attached test fittings.
The allowable production leakage is 0.1 SCFH. While reinstalling the valve topworks, a test circuit used for original plant start-up was found connected to the valve. Between August 3 and the 30th, the valve was opened satisfactorily 10 times with the test loop intermittently connected.
On October 7,1980, the A-valve open&d at a reactor pressure of 1025 psig, and remained open until the vessel depressurized.
Investigation showed that nitrogen pressure in the actuating line was 160-165 psig, which caused leakage in the solenoid valve.
This pressurized the actuator diaphragm, opened the 2nd stage, and caused the main disk to open. The design of the solenoid valve is such that pressures in excess of 135 psig prevent the valve from reclosing. Corrective measures.were to maintain less than 135 psig in the supply line and monitor this by surveillance once per shift.
EDITDN EDIEDN COMPANY Mr. Don.enic B. Vassallo, Chief Februa ry 11, 1982 Page 3 On October 31, 1980 the A-valve opened again and remained open until reactor pressure fell to 20 psig. The nitrogen supply pressure in the actuating line was again found to be 165 psig. A regulator, which had been placed in parallel with the existing regulator in August of 1980, had frozen and failed open, adding excessive pressure within the line.
Corrective measures were to take the nitrogen line out of service, and to feed the solenoid valves from station instrument air regulated to 110 psig. This operational method was continued until the 1981 outage.
In November,1981,during the Refueling Outage, B & C valves were tested at Wyle Laboratories for recalibration of set points. They actuated at 1136 psig and 1230 psig on first pop.
Leakage by the pilot stage was excessive and,on disassembly, the pilot disks were found to be badly eroded and covered with a tightly adhering black deposit.
The A and D valves were then tested. The A valve had a measured leakage of 880 lbs./hr. by the pilot seat, and actuated at 1210 psig on the second attempt to pop.
Its pilot disk was also found to be badly eroded and covered with a black deposit.. The D valve had leakage of 8.5 lbs./hr., and it actuated at 1136 psig.
The disk showed incipient erosion and had a slight black deposit. Also, valves A, B, and C leaked considerably, and the D val te leaked slightly at the body to bonnet flanges.
During the inspection of valve D, an 8 inch long crack was discovered in the body cavity behind the disk piston cylinder wall.
This crack is considered irreparable and is discussed later in this response.
After these tests, the logs of tail pipe temperatures were procured from the computer; they indicated that A and C valves had leaked from start-up in May 1980, the B valve from September 1980, and the D valve from August 1981.
An additional test was conducted at Pilgrim to determine if the solenoid valves were suceptible to leakage due to overpressurization.
The results are contained in the table below:
Val ve PSIG at which excessive leakage beaan A
150 B
155 C
165 D
195 Causes of Failure The cause of failure of the valves to maintain set point was a change in the design of the downward spring and the upwa,rd pressure forces it encountered.
Excessive pilot and body to bonnet leakage, and lower than-design body and bonnet temperatures caused this problem. We are investigating how much of the set point drift is attributable to each of the above causes.
Our initial belief is that the cause of the failure to actuate (open) in the relief mode within specified ranges of pressure was seat leakage, and body to bonnet leakage caused by an inadequate gasket design.
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i PO3 TON EDlHON COMPANY Mr..Domenic B. Vassallo, Chief February 11, 1982 Page 4
.0nce seat leakage begins, the seats degrade continuously. leading to consistent increases in leakage. Leakage by valve seats-is controlled by differential pressure across the seats, the conformance of the surfaces of the seat and disk, and the contact stress between seat and disk.
The variation in pressure across the seats of valves at different plants is insianificant. Also, since the disks are lapped onto the disks by, or under the supervision of, the manufacturer, the surface conformance is not i
l expected to vary significantly between valves in stations..The contact stress between seat and disk does vary significantly between stations or between valves in the same station. 'The contact stress depends on the net force i
acting on the disk, which varies directly as the difference between set pressure and operating pressure (the simmer margin) and inversely as the contact area i
between seat and disk.
GE recommends a simmer margin of 120 psi.
Pilgrim operated at approximately 50 psi, while other stations operate at higher margins. Therefore, the Pilgrim valves were somewhat more likely to simmer or chatter slightly under line vibration and slight pressure surges and oscillations than those of other stations.
It is also probable that the excessiv'e nitrogen pressure condition, which in October 1980 caused the A valve to lift, was sporadically occuring from the time the valves were initially put into service in May 1980.
This would have partially pressurized the air operator diaphragm and further reduced the simmer margin, increasing the probability of seat leakage.
Actions to Assure Design Performance
. e believe that by taking the following actions the valves can be put back into Wservice, and that they will maintain the same safety margins as were originally considered in the design of the plant:
i (A) Actions Necessary to Restore Valve Operability (1) All eroded disks and seats and other damaged parts must be repaired or replaced.
(2) The body to bonnet gasket joint must be altered to take a gasket which has been proven to be effective in operation.
(3) The reconditioned valves must be reset on steam.
All these actions have been completed.
(B) Actions to Maintain Opertbility (1) The leakage rate past the pilot stage must be monitored, and the valve repaired before thoseleakage conditions are reached which cause excessive set point drift.
A possible method of monitoring the leakage rate is 'to correlate tailpipe temperatures to leakage. Boston Edison has. requested such a correlation from G.E. - However, preliminary attempts by G.E. at developing this cor-l relation indicates that such a correlation may be impossible to develop.
Therefore, to satisfy this concern, Boston Edison shall propose a l
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CD: TON Eol:ON COMPANY ltr. Domenic B. Vassallo, Chief February 11, 1982 Page 5 Technical Specification that requires the reporting to the NRC of 0
tailpipe temperatures in excess of 212 F.
This reporting Technical Specification is an interim measure, and will be replaced with a more specific Technical Speciff cation if an accurate tailpipe / leakage correlation is developed.
(2) A critical leakage rate must be developed by GE or Target Rock analysis and/or testing to confirm leakage conditions which would be acceptable.
(3) A determination must be made as to whether body to bonnet leakage significantly contributed to set point drift.
If'it did, then an additional safety analysis must be performed to judge whether temoer-ature monitoring of the body / bonnet is necessary until the new joint has been proven dependable.
(C) Actions to Inprove Reliability (1) The feasibility of optimizing the simmer margin will be investigated together with an evaluation of the consequences of so doing.
(2) The nitrogen supply system will be modified by installing relief valves and pressure control valves of greater reliability, thereby allowing the system pressure to be held more reliably at 110-120 psig. These modifications will also include an alann that will alert the operators of potential system overpressure at 115 psig. A full port relief valve will actuate at 120 psig.
Section II:
Cracked Body on SRV "D" Purpose We submit this portion of the response to demonstrate that Target Rock S/R Valves can be used with reasonable confidence of not having an undiscovered crack in their bodies.
Background
During the recertification testing of the SRV's at Wyle in November 1981, a crack was discovered in the body cavity under the two stage bonnet and behind the main piston cylinder wall in Valve D, Base #10. The crack was about 8" long, and plainly visible to the naked eye.
It extended from just under the shoulder supporting the cylinder wall and axially along the body, to the top flange area of the cavity. The crack was between two weld repairs and there were a total
.of about 5 weld repairs in the cavity, all extending in the axial direction.
- - A.short exploratory evacuation of the crack showed its depth to be greater than Finch.
The valve was taken out of service and is considered irreparable.
BOLTON EDIHON CO M PANY
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Mr. Domenic-B. Vassallo, Chief February 11, 1982 Page 6 Cause In 1979, after the body was remachined for the two-stage conversion, the seat region just below the location of the present crack was the subject of a PT examination. We believe a crack of the size and appearance of the one discovered could not have escaped notice during this examination.
Therefore, we believe that the crack surfaced sometime during the last operational cycle. The most probable explanation of this crack is that it originated from a subsurface crsting flaw, or a region of high residual stress caused by the repair welds, and was driven to the surface by 'ressure, themal, and operational stress
. cycling.
Acceptability of Valves for Continued Operation Basec' on the inspection of the other valves, and on QC examinations of RT records, we have no reason to believe that a generic defect exists, or existed, to cause the discovered crack. The failed valve is part of the deviant statistic.
Therefore, unless engineering investigations that are currently in progress demonstrate otherwise, the valves can be considered to be in conformance to the original design specifications, and that continued operation using repaired and spare valves is justified.
Section III:
Impact of Higher SRV Setpoints During Cycle 5 Effect of Higher S/RV Setpoints on Abnormal Transients - Peak Overpressure and MCPR Limits For peak overpressure analysis, the limiting event is MSIV closure with flux scram.
For MCPR analysis, the limiting event is Load Rejection Without Bypass (LRNBP).
Analyses of these events were perfomed (with standard reload methods) and results for the higher setpoints are compared with the Pilgrim Cycle 5 reload analysis (NED0-24224). The analysis results show that peak overpressure would be within the ASME code limit, and that there would be no impact on MCPR if either of these evcats had occurred during Cycle 5.
Table 1 shows that for the worst case MSIV closure event, the peak overpressure increases by 25 psi relative to the reload analysis, but is still within the ASME code limit of 1375 psig.
For the LRNBP event, a similar sensitivity study was performed which showed that the peak neutron flux and surface heat flux occurred prior to relief valve opening for both the original and the as-found setpoints. Therefore, there would be no impact on MCPR due to higher setpoints.
ED3 TON EoirON COMPANY Mr. Domeric B. Vassallo, Chief -
February 11, 1982
' Page 7
. TABLE 1 PEAK OVERPRESSURE i
P, Peak Vessel Pressure, (psig).
,PSL, Peak Steam Line Pressure, (psig)
, y Higher Setpoints
-Higher Setpoints Reload Analysis Analysis Reload Analysis Analysis 1328 1352 1341 1366 1
i Effect of Higher SRV Setpoints on Peak Cladding Temperature (PCT) for the LOCA Event i
The effect of.the as-found-setpoint deviations on the PCT for-LOCA(s) was
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evaluated and analyzed by using the approved 10CFR50 Appendix K models.
4 During intermediate and large pipe break LOCA's, the SRV's would not actuate since the quick ~ inventory loss-causes a decreasing vessel pressure from scram, and the SRV setpoints are never reached.
I However, vessel pressure following a scram _ during a small break accident tends to' remain high. This is because the break flow is insufficient to depressurize the vessel, therefore steam generation continues in the core.due to decay heat.
Therefore, the SRV setpoints are expected to be reached, and the effect of in-creased SRV setpoints was investigated for small-break LOCA's..
I was' determined to be a 0.05 ftgus break sizes, the most limiting in terms of PCT
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From previous analysis of vari recirculation suction line break with an assumed failure of the'HPCI system.
For this case, the SRV;s were modeled using the "as-found" setpoints. The two valves with as-found setpoints of 1210 and 1230 psig are not significant, since the vessel pressure will not exceed the opening of the lowest setpoint valves (1136 psig). This SRV setpoint increase was found to cause approximately a 19 F increase in PCT, which generates a PCT well below the 2200*F limit.
It is concluded that the higher SRV setpoints have no significant impact on PCT's for LOCA's.
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Effect of Higher Setpoints on Capability of the RCIC System
~ The design objective of the reactor core isolation cooling (RCIC) system is to provide sufficient water to cool the core during reactor' isolation when-the nonnal reactor heat sink, the main condenser, and feedwater makeup flow are : unavailable. After isolation, the SRV's cycle to maintain vessel pressure within acceptable limits, and the RCIC system is automatically. put into
' operation by a low water level signal. Since the RCIC system design specification only requires constant rated flow up to a vessel pressure of 1135 psia, increased SRV setpoints above 1135 psia could compromise RCIC performance. However, upon evaluation of the RCIC pump / system characteristics, it was determined
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COOTON EDIEON COMPANY Mr. Domenic B. Vassallo, Chief February 11, 1982 Page 8 that the existing RCIC system is capable of providing 100% rated flow at or above 1136 psig. This leads to the conclusion that there will be no degrad-ation of the RCIC system performance due to the "as found" deviation in the SRV setpoints.
Although no impact on RCIC performance was found, an analysis was performed assuming loss of feedwater, reactor scram on low water level, and reactor isolation on low low water level concurrent with the RCIC initiation signal. The RCIC flow rate was reduced to demonstrate the margin available in pumping capacity.
The results of this analysis demonstrate that even under this postulated most limiting condition for RCIC operation, the minimum water level reacned remains more than 4.5 ft. above the top of the active fuel.
Hence, there is no concern about RCIC performance with the "as-found" setpoints.
Combined Effect of Water Level Instrumentation Error and Higher S/RV Setpoints on PCT's for LOCA's In Boston Edison's report " Safety Evaluation for the Drywell Event,"
transmitted to the NRC by BECo letter 82-27, a bounding.value for the level instrumentation error of 10 inches was used to conservatively show the effect of delayed ECCS initiation. A combination of this 10 inch error in the water level instrumentation and the raised SRV setpoints was evaluated with respect to PCT's during small break LOCA's.
The postulated PNPS worst case of a 0.05 ft2 break was reanalyzed, taking into account both the reduced water level and increased SRV setpoints.
This leads to a delay in ADS actuation of 12 seconds with a corresponding 49*F increase in PCT, but the PCT was still well below the 2200 F limit.
Therefore, the combined effects are negligible in terms of PCT's for LOCA's.
Summary of Section III Boston Edison believes that the various analyses performed, the results of which are presented above, demonstrate that adequate overpressure protection existed during Cycle 5.
Section IV: Technical Specification As stated in Section I of this subm'ittal, Boston Edison will submit a proposed Technical Specification that requires the reporting of tailpipe temperatures that exceed a specified parameter. This Technical Specification will be the subject of a separate submittal. We are developing this Technical Specification
'in. conformance with information developed at.a meeting between NRC and General Electric conducted on January 15, 1982, and reaffirmed in a telephone conversation
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between NRC and Boston Edison on January 20, 1982.
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CO3 TON EDIEON COMPANY Mr. Domenic B. Vassallo, Chief February 11, 1982 Page 9 Boston E'Jison believes the above information satisfactorily addresses the concerns contained in (Reference (b). We also believe that all the analyses, inspections and testing that have been performed pertaining to this issue indicate that it should not constrain Pilgrim's resumption of power operation. Should you require further_ information concerning this issue, please contact us.
Very truly yours,
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} w W. H. Deacon, Acting Manager Nuclear Operations Support Dpet.
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