ML20059N222
| ML20059N222 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 09/26/1990 |
| From: | Owen T DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9010110007 | |
| Download: ML20059N222 (7) | |
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, truke itver Compaqy (80?)S31 3000 {
Catauh KuclearStation 1:0 lha 2%
Cimer, iC29710 DUKEPOWER September 26, 1990 Document Control Desk U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Subject:
Catawba Nuclear Station Docket 50-413 and 50-414 C90-021-1, Rev 1 ,
1-C90-0075, Rev 1: 2-C90-0248 f Gentlemen: ,
Attached is a report on SETPOINT DRIFT ON PRESSURIZER SAFETY VALVES DURING SUCCEESIVE SURVEILLANCE TESTS. This report'. is being submittod as a "Special" Report with potential 10CFR Part 21 reportability concerns.
Very truly yours,
'l y Tony . Owen ,
Station Manager ken: REPORT.SP xc: Mr. S. D. Ebneter American Nuclear Insure ~s ;
Regional Administrator, Region II c/o Dottie Sherman,^ANI Library. ;
U. S. Nuclear Regulator Commission The Exchange, Suite 74%
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101 Marietta Street, NW, Suite 2900 270 Farmington Avenue Atlanta, GA 30323 Farmington, CT 06032 t i
M & M Nuclear Consultants Mr. K. Jabbour s 1221 Avenues of the Americas U. S. Nuclear Regulatory Commission New York, NY 10020 Office of Nuclear Reactor Regulation Washington, D. C. 20555 INPO Records Center Suite 1500 Mr. W. T. Orders 1100 Circle 75 Parkway NRC Resident Inspector Atlanta, GA 30339 Catawba Nuclear Station *
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I DUKE POWER COMPANY r CATAWBA NUCLEAR STATION j PROBLEM INVESTIGATION REPORT NO. 1-C90-0075, Rev. 'l
- SETPOINT DRIFT ON PRESSURIZER SAFETY VALVES ~ .
DURING SUCCESSIVE SURVEILLANCE TESTS ABSTRACT -
On February 3-4, 1990, two of'the thhee Pressurizer Safety Valves (PSVs) were -
removed from the Unit 1 Pressurizer to meet the Techt leal Specification' (T/S) -l Surveil. lance test ;guirement. Both valves were shi., ped to the test facility at Wyle Laboratories. On February 23, Maintenance Engineering Services (MES) , l personnel were contacted that one valve-failed to neet the 2485 psig +/ - 1% .
setpoint requirement'during its as-received setpcint test. Unit I was in "No.
Mode",' core defueled, at this time. . Additionally, on June 22, 1990, two of the. 3 three PSVs were. removed from the Unit 2 Pressurizer.and shipped to Wyle- ,
Laboratories. On July 22, MES personnel were contacted that both valves failed to meet the 2485 psig +/- 1% setpoint requirement. Unit 2 was in "No Mode" at i this time. A review of past surveillance tests indicated that the PSVs have i failed to meet the setpoint requirement in 60% of their as-received setpoint L tests. Setpoint drift on the PSVs beyond the T/S allowed +/- 1% acceptable range is attributed to a (manufacturer) Functional Design' Deficiency. The cause
- of the setpoint drift is attributed to the affect normal plant operation and system condition can have on PSV performance. A Safety Evaluation will be performed, and station documents will be revised, as needed, 'o evaluate increasing the acceptable range to +/- 3% on the PSV setting. . This report is '
being submitted as a special Report wi'h potential 10CFR Part 21 reportability concerns.
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.. ., DUKE POWER COMPANY / CATAWBA NUCLEAR' STATION <
t i .'P*.R 1-C90-0075/Specicl Report,.R:v. 1 Pag) 2 - i
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t BACKGROUND !
The Reactor Coolant [EIIS:AB) (NC) System consists of four heat transfer loops connected in pr.rallel to the Reactor Vessel [EIIS:VSL]. Each loop contains a Reactor Coolant Pump [EIIS:P] and a Steam Generator [EIIS:HX] (S/G). The B loop' i alscs includes a Pressurizer, a Pressurizer Relief Tank (PRT), interconnecting .{
piping [EIIS: PSP] and instrumentation necessary for operational control, j NC System pressure is controlled by the use of the Pressurizer where water and a steam are maintained in equilibrium by electric heaters [EIIS:EHTR] and water sprays. -Steam can be formed (by the heaters) or condensed (by the Pressurizer-spray) to reduce pressure variations due to contraction and expansion of-the !
Reactor coolant. Three spring loaded safety valves-[EIIS:V] 1(2)NC-1, 2.and 3, ;
are connected to the Pressurizer and discharge to the Pressurizer Relief. Tank (PRT).
The three Pressurizer Safety Valves (PSVs) per Unit are of the totally enclosed l $
pop-type. The valves are manufactured by Dresser,.Model 6-31749A-2-XNC019,'and are spring-loaded, self-activated with back pressure compensation. The combined i capacity of the valves is equal to, or greater than, the maximum surge rate resulting from complete loss of load without Reactor Trip or any other control. 'i Temperature indicators [EIIS:XI].in the safety valve discharge' manifold alert :
the Operator to the passage of steam due either to leakage or valves lifting.
Each Unit's Pressurizer is equipped with three Power Operated Relief Valves (PORVs) which limit system pressure for a large power mismatch and thus prevent actuation of the fixed high-pressure Reactor trip. The PORVs are operated automatically or by remote manual control. The operation of these valves also limits the undesirable opening of the spring-loaded safety valves. Remotely I operated valves are provided to isolate the inlet to each PORV.if-excessive '
l leakage occurs.
Each Unit's PRT condenses and cools the discharge from the PSVs and PORVs.
Steam is discharged through a sparger pipe under the water level. The PRT is l equipped with an internal spray and a drain which are used to cool the tank following a discharge. 'The PRT is protected against a discharge exceeding the design value by two rupture discs which discharge into the Reactor Containment.
Technical Specification (T/S) 3.4.2.1 requires that a minimum of one PSV be operable with a lift setting of 2485 psig +/- 1% for Unit operation in Mode.4, l Ilot Shutdown, or Mode 5, Cold Shutdown. With no PSV operable, inuediate suspension of all operations involving positive reactivity changes and placement ;
of an operable residual heat removal loop into operation is required.
5 T/S 3.4.2.2 requires all PSVs be operable with a lift. setting.of 2485 psig +/- '
1% for Unit operation in Mode 1, Power Operation, Mode'2, Startup, and Mode 3.
Hot Standby. With one PSV inoperable, action shall be taken to restore the l .
I inoperable PSV to operable status within 15 minutes, or be in at least Mode 3.. -
within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least Mode 4 within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
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DUKE POWER COMPANY / CATAWBA NUCLEAR STATION
- PIR 1-C90-0075/Specicl R: port, R;v. 1 l
- P g3 3 T/S 4.0.5 identifies that inservice inspection and testing of the PSVs shall be performed in accordance with surveillance requirements identified in the ASME Boiler and Pressure Vessel Code,Section XI, 1983 Edition including Addenda through the Summer 1983 Addenda (applicable Edition and Addenda required by 10CFR part 50, Section 50.55a(g)). Subsection IWV-3510 of the ASME Code identifies the testing schedule which requires all PSVs be tested once every-five years. Subsection IWV-3510 required that setpoint testing be in accordance with ASME PTC 25.3-1976, Safety and Relief Valves performance Test Codes.
The 1986 Edition of the ASME Section XI Code,7 Subsection INV-3510, references that setpoint testing be in accordance with ANSI /ASME OM-1-1981, Requirements '!
For Inservice Performance Testing of Nuclear Power Plant Pressure Relief i Devices. This document allows a +/- 3% acceptable range for setpoint test.
Surveillance testing to subsequent edition of codes and addenda, or portions thereof, is acceptable to 10CFR Part 50, Section 50.55a(g)(3)(v). . However,.the present T/S limitation require that the +/- 1% acceptance range be maintained at Catawba.
EVENT DESCRIPTION On February 3-4, 1990, PSVs INC-001 and 1NC-002, Serial Numbers (S/Ns) BS-02867 .
and BS-02872, were removed from the Unit 1 Pressurizer to be tested per the T/S l Surveillance requirements (reference Work Requests 4401. SWR and 4402 SWR).
Replacement valves S/Ns BS-02870 and BS-02871, previously tested and verified to be at the required setpoint, were installed. On February 10, the valves removed-l j were transported to the test 'acility at Wyle Laboratories. On February 23, l Maintenance Engineering Servis.es (HES) personnel were contacted that valve S/N l BS-02867 (INC-001) failed to meet the 2485 psig +/- 1% requirement during its as-received setpoint test. Unit I was operating in "No Mode", core defueled, at ,
this time.
Station Compliance and MES personnel discussed the concerns of PSVs setpoints drifting outside of the T/S allowed tolerance. An MES review of past setpoint test results indicated that of the 13 tests performed on the PSVs, there were 7 valves that failed the Wyle Laboratories as-received setpoint test.' MES "
initiated Problem Investigation Report (PIR) 1-C90-0075 identifying the recent drift in setpoint of valve S/N BS-02867 as a continuing problem.
- Valve S/Ns BS-02867 and BS-02872 were refurbished, and final setpoint testing -
was acceptable. These valve were returned to Catawba as replacement spares.
On June 22, 1990, PSVs 2NC-001 and 2HC-002, S/Ns BS-02868 and BS-02866, were removed from the Unit 2 Pressurizer to be tested per T/S (reference Work Requests 4472 SWR and 4499 SWR). The valves removed were then transported to Wyle Laboratories for a setpoint verification test. On July 22, MES personnel were contacted that both valves failed to meet the 2485 psig +/- 1% requirement
.during their as-received setpoint test. Unit 2 was operscing in "No Mode" at this time. Valve S/Ns BS-02867 and BS-02872 were instal.cd as replacements.
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. DUKE POWER COMPANY / CATAWBA NUCLEAR STATION
- PIR 1-C90-0075/Special R: port, Rev. 1 l
- p ge 4 CONCLUSION This incident is attributed to a (manufacturer) Functional Design Deficiency due to.the PSVs not maintaining the T/S setpoint of 2485 psig within the +/- 1%
acceptable range. Catawba has experienced a drift in setpoints outside of the
+/- 1% tolerance in 60% of the PSVs tested at Wyle Laboratories (9 failures of 15 tests performed). A search of the Nuclear Plant Reliability Database System (NPRDS) identified 62 failures of PSVs to be within their setpoint requirements -
during surveillance testing (valves tested in response in equipment problems, t
1.e. seat leakage, were'not included). Of these 62 failures, which.does not include any unreported incidents, 48 involved valves manufactured.by Crosby valve and Gauge and 14 involved valves manufactured by Dresser Industries.
Infor:aation on 11 of 23 valves indicated that they had drifted more than +/ .3%
(the remaining 39 valves did not identify the degree of setpoint drift). The cause of setpoint test failure was identified as undetermine6 or as normal setpoint drift due to aging and thermal cycling in 55 of the 62 failures.
Improper test methods and procedures resulted in 7 of the failures. The corrective action taken in 68% of the failures was to adjust the.setpoint within ;
the acceptable range with no additional actions required. 'In 21% of the failed valves, disassembly, inspection, and cleaning did not identify a cause other than normal setpoint drift. The failure history of the PSVs tested for Catawba is consistent with that of the NPRDS data. There have been 9 tests failing to meet the +/- 1% acceptance requirement at Catawba. Of these 9, only1 was :
greater than +/- 3% (at 3.6% low). '
The ability of safety valves to repetitively and consistently perform within the
+/- 1% Code requirement is recognized as an industry wide concern. Recent developments include the effects of loop seal on the ability of safety valve to ,
consistently repeat lifts within the desired setpoint range. -The Westinghouse Owners Group is presently pursuing discussions to address the loop seal concerns >
in an effort to develop test methods and guidelines to improve setpoint repeatability. Catawba does not utilize loop seals in the psi application and therefore does not have loop seal concerns. Other developments include test I
l studies of safety valves to reseat at a desired pressure below:setpoint without an extended blowdown period (EPRI Test Report NP-2770-LD and NP-2628-LD).
Results from this study have been used to develop adjustments and tighter l_ contro.ls on the PSV ring settings. The affects of_an extended blowdown on the
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PSVs are presently being evaluated for the Catawba feedwater line break accident (LER 413/90-001).
The utility industry's input was solicited using the NUCLEAR NETWORK and the NPRDS system. From the response that was obtained.(10 operating plants), it can be concluded that pSVs are not meeting their as-found +/- 1% setpoint acceptance-requirement at a 42% failure rate. The variance in the information obtained is '
large with some stations having little trouble _ meeting acceptance requirements.
! to stations experiencing a'75% failure rate. From this information, it is concluded that there are generic issues that need to be addressed concerning pSV setpoint drift.
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. . 9 UKE POWER COMPANY / CATAWBA NUCLEAR STATION
+ P D.1-C90-0075/Specicl R: port, R;v. 1 l
' Page 5
,.t Catawba, PIR 0-C90-0026 has been issued to address concerns of setpoint drif co the Main Steam Safety Valves (MSSVs). These concerns were originated when the valve manufacturer changed the equipment calibration constants that effected the tolerance by 1.0% to 1.5% of the original setting, which have a ,+ 1% ,
acceptance tolerance of setpoint. PIR 0-C90-0026 identified that setpoint drifts greater than the +/- 1% experienced on 55% of these tests, indicating that the MSSVs and the PSVs have similer setpoint drift concerns.
The ability of the PSVs (or MSSVs) to consistently repeat setpoint lifts at any given pressure can be affected by many uncontrollable variables, including changes in temperatures, pipe loads and stresses, rate of pressure increase, condensate, and seat leakage. These variables may affect the relief setting in that the test conditions and the actual conditions may not be the same. Also, 'i actual conditions may in themselves vary enough to causo variances in lift settings. Wyle Laboratories utilizes special tests end control procedures and methods to simulate the actual conditions that are err ~'ted to exist et Catawba (Wyle Test Procedure 1028). These efforts have contributed to maintaining relief settings consistently within +/- 3% of setpoint during successive surveillance testing. The single test that was outside of +/- 3% is considered to be an anomaly and is not expected to be recurring. Wyle Laboratories will refurbish this valve and inform MES of any observed deficiencies that may have contributed to its setpoint variance. A Safety Evaluation;will be performed to evaluate the effect of relief setting varying +/- 3% of setpoint dua to the ,
uncontrollable variables that affect valve performance. These efforts will be I to evaluate the acceptability of as-found (Wyle Laboratories as-received) tests '
varying +/- 3% of setpoint. However, it is expected that the as-left (Wyle Laboratories as-shipped) test settings will remain within +/- 1% of setpoint.
This will provide a 2% buffer, on both the high and low side, for setpoint drift without involvin'; an operability question. Thisposition,if.deterkned acceptable, should be documented in the T/Ss, T/S Bases, or T/S Interpretations for Safety /Rollei /alves.
l The inability of safety valves to meet performance of the +/- 1%-test 4 requirement is identified as a recurring problem. The roccmmended and in-progress evaluation and the Westinghousd Owners Group offorts is expected to address these concerns and implement changes that will improve safety valve performance. A review of the Operating Experience Program (OEP) database '
revealed no previous events where equipment setpoints varied greater than their i designed tolerance. However, the two incidents identified in the report meet the Nuclear Safety Assurance detinition of a recurring problem.
This report is being submitted as a Special Report in that setpoint drift
- greater than the +/- 1% required setting range could occur in subsequent '
as-found surveillance tests. The cause of the setpoint drift could not be attributed to a known condition that would recult in the inopelanility of the .
PSVs. The setpoint drift is attributed to the effect that normr'. plant operation and system conditions can have on the ability of the PSV to consistently relief at the required setting. Therefore, no T/S violat. ion has occurred. #
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DUKE. POWER-COMPANY / CATAWBA NUCLEAR STATION
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f PIR 1-C90-0075/Specici R: port, R;v. 1
' Pag 6 6 This report is being submitted as potentially 10CFR Part 21 reportable; the PSV is considered a " basic component"; a drift.in setpoint greater than +/- 1% is considered a " defect"; and the existence of a." substantial safety hazard" is dependent upon the affect this retpoint drift.can have on protecting the Reactor Coolant System for the individual' plant normal operation and analyzed accident conditions. The requirement that the PSVs function within +/- 1% of setpoint-was identified within manufacturer purchase specification CNS 1205.09-00-0001, Pressurizer and Main Steam Safety. Valves. .'Setpoint drifts outside +/- 1% of-setpoint due to normal operating conditions is in violation of the purchase specification.
CORRECTIVE ACTION SUBSEQUENT
- 1) An NPRDS search was performed to identify an industry wide concern on setpoint drifts in PSVs.
PhANNED
- 1) A Safety Evaluation will be performed to address setpoints drifts of
+/- 3% as an acceptable condition on the PSVs.
- 2) Station Documents (T/Ss, T/S Bases, or T/S Interpretations) will be revised / developed to incorporate the results of the Safety Evaluation.
- 3) The Safety Analysis of this report will be revised to incorporate the results of the safety evaluation.
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- 4) For the >3% test results, this valve will be inspected by Wyle Laboratories for any abnormalitica that could have affected the.
setpoint.
SAFETY ANALYSIS l
Variances in the PSV's setting. greater than +/- 1% of setpoint have not been considered in the present safety analysis assumptions. The PSVs are required to
' function to limit NC System pressure during incidents invol.ving a decrease in . q heat removal by the secondary system, decrease in Reactor coolant flow rate, and anomalies in reactivity and power distribution. A safety evaluation will be- ,
performed to address variances in PSV settings to +/- 3% of setpoint. The results of the safety evaluation will be incorporated into the. Safety Analysis.
of this report.
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