ML20034A079
| ML20034A079 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 04/04/1990 |
| From: | Owen T DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| REF-PT21-90, REF-PT21-90-045-000 1-C90-0075, 1-C90-75, PT21-90-045-000, PT21-90-45, NUDOCS 9004190280 | |
| Download: ML20034A079 (7) | |
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Clom,S C :97l0 DUKEPOWER April 4, 1990' Document Control Desk U. S. Nuc1 car' Regulatory Commission Washington, D..C.
20555
Subject:
. Catawba Nucleart Station Docket No. 50-413 PIR 1-C90-0075; IIR C90-021-1 Gent 1cmen:
Attached-is Problem Investigation Report 1-C90-0075,: submitted iconcerning SETPOINT DRIFT'ON-PRESSURIZER SAFETY" VALVES DURING SUCCESSIVE SURVEILLANCE TESTS.
This report is being submitted as a Special Report with potential 10CFR Part 21~reportability-concerns.
This event was-considered to be of:no significance with respect to the-health and safety of-the public.
Very truly yours,
\\
b Tony
. Owen Station Manager kob\\ REPORT.SP xc:
Mr. S. D. Ebnetor American Nuclear Insurers-Regional Administrator, Region II c/o Dottie Sherman', ANI Library U. S. Nuclear Regulator Commission The Exchange, Suite 245-101 Marietta Street, NW, Suite 2900 270 Farmington Avenue Atlanta, GA-30323 Farmington,'CT 06032
'M & M Nuclear Consultants Mr. K. Jabbour 1221 hvenues 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
.lhtlanta,_GA 130339 Catawba Nuclear Station Y.E n.
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9004190280 900404-DPDR. ADOCK 05000413 l
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b6 DUKE POWER COMPANY CATAWBA NUCLEAR STATION-PROBLEM INVESTIGATION REPORT NO. 1-C90-0075 SETPOINT ORIFT ON PRESSURIZER SAFETY VALVES DURING SUCCESSIVE SURVEILLANCE TESTS ABSTRACT On February 3-4, 1990, two of the three Pressurizer Safety Valves (PSVs) were removed from the Pressurizer to meet the Technical Specification (T/S)
Surveillance, test requirement.
Both. valves;were shipped to the: test facility at-Wyle Laboratories. On February 23, Maintenance Engineering Services (MES) personnel were contacted that one valve failed to meet-the 2485 psig +/- 1%
H setpoint requirement during its as-received:setpoint test. Unit 1 was in "No Mode", core defueled, at this time.
Further review by MES personnel of past surveillance tests indicated.that the PSVs'have failed to meet the setpoint requirement in 53.8% of their as-received setpoint tests.
Setpoint drif t 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 i
will be revised,.as needed, to evaluate increasing the acceptable range to +/-
3% on~the PSV setting. This report is being submitted as a Special Report with potential 10CFR Part 21 reportability concerns.
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-~ DUKE POWER COMPANY / CATAWBA NUCLEAR. STATION PIR 1-090-0075/Special Report Page 2 b
BACKGROUND The Reactor Coolant [EIIS:AB] (NC) System consists of four heat transfer loops connected in parallel 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 also includes a Pressurizer, a Pressurizer Relief Tank (PRT), interconnecting '.
piping [EIIS: PSP] and instrumentation necessary for operational control-.
NC System pressure is controlled by the use of the Pressurizer where water and 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 i
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.
3 The three Pressurizer Safety Valves (PSVs) are of the totally enclosed 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 capacity of the valves is equal to, or greater than, the maximum surge rate resulting from complete loss of load without Reactor Trip orf any other control.
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'.
.The Pressurizer is equipped with three Power. Operated Relief Valves (PORVs) which limit system pressure for a large power mismatch and thus1 prevent actuation of the fixed high-pressure Reactor trip.
The PORVs aro' operated automatically or by reste manual control.
The operation of these valves also limits the undesirable opening of-the spring-loaded safety valves.
Remotely operated valves are provided to isolate the' inlet to each PORV if excessive leakage occurs.
The PRT condenses and cools the discharge from the PSVs and PORVs.
Steam is
.l discharged through a sparger pipe under the water level.
The PRT is equipped-i 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% during operation in Mode 4, Hot Shutdown, or Mode 5, Cold Shutdown.
With no PSV operable, immediate suspension of all operations involving positive reactivity changes and placement of an operable residual heat removal loop into operation is' required.
T/S 3.4.2.2 requires all PSVs be operable with a lift setting of 2485 psig +/-
1% during 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 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 />.
( DUKE POWER COMPANY / CATAWBA-NUCLEAR STATION-PIR 1-C90-0075/Special.R: port
.Page -- 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 Coda' identifies the testing schedule which requires all PSVs be tested once every J
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, Subsection IWV-3510,. references that setpoint testing _be in accordance with AMSI/ASME_0M-1-1981, Requirements For Inservice Performance Testing of Nuclear ~ Power Plant: Pressure Relief Devices.
This document allows a +/- 3% acceptable range for setpoint test.
Surveilance testing to subsequent edition of codes and addenda, or portions thereof, is acceptable to 10CFR Part 50, Section 50.55a(g)(3)(v). However, present T/S limitation required that'the +/- 1% acceptance range be maintained at Catawba.
EVENT DESCRIPTION On February 3-4, 1990, PSVs 1NC-001'and 1NC-002, Serial Numbers (S/Ns) BS-02867 i
and BS-02872, were removed from the Pressurizer to be tested per the T/S-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 were transported to the test facility.at Wyle Laboratories..On February 23,-
-Maintenance Engineering Services (MES) personnel,were contacted that valve S/N 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 l
test results indicated that of the 13 tests performed on the PSVs, there were 7 l
valves (53.8%) that failed the Wyle Laboratories as-received setpoint test.
These, test indicated variances in setpoints from 2.5% high (2546 psig) to l'.5%
low (2448 psig). MES initiated Problem Investigation Report (PIR) 1-C90-0075 l-identifying the recent drift in setpoint of valve S/N BS-02867 as a' continuing l
problem.
CONCLUSION a
T'eis 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 53.8% of the PSVs tested at Wyle Laboratories.
A search of the Nuclear Plant Reliability Database System (NPRDS) identified 62 failures of PSVs to be within their setpoint requirements during surveillance testing
- ' DUKE POWER COMPANY / CATAWBA NUCLEAR STATION
' PIR 1-C90-0075/Special Repert-
.Page 4:
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h (valves tested in response in equipment problems, i.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.
Information on 11 of 23
. valves indicated that they had drifted more than +/- 3% (the remaining 39 valves did not identify the degree of setpoint dri_ft). _The cause of setpoint test failure was identified as. undetermined or as normal setpoint drift due to-aging:
and thermal cycling in-89%'of the_62: failures ~
Improper test methods and-procedures resulted in 11% 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 any results that would indicate a.
cause other than normal setpoint drift. The failure history of the:PSVs tested for Catawba =is consistent with that of the NPRDS data, except that the Catawba PSVs have not drifted more than +/- 3% of the T/S setting.
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 4
in an effort to develop test methods and guidelines to improve setpoint repeatability. ' Catawba does not utilize loop seals in the PSV application =and therefore-does not have loop seal concerns. Other developments include _ test 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 controls on the PSV ring settings. The affects of an. extended blowdown onithe i
PSVs are presently being evaluated for the Catawba feedwater line_ break accident (LER 413/90-001).
At Catawba, PIR 0-C90-0026 has been issued to address concerns of setpoint drift
.on 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%
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acceptance tolerance of setpoint.
PIR 0-C90-0026 identified that setpoint drifts' greater than the +/- 1% experienced _on 55% of these tests, indicating l
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that the MSSVs and the PSVs have similar 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 l
changes in-temperatures, pipe loads and stresses, rate of pressure' increase, I
condensate, and seat leakage.
These variables may affect-the relief setting in L
that the test conditions and the actual conditions may not be the same. Also, actual conditions may in themselves vary enough to cause variances in lift settings. Wyle Laboratories utilizes special tests and control procedures and l
methods to simulate the actual conditions that'are expected to exist at' Catawba (Wyle Test Procedure 1028).
These efforts have contributed to maintaining.
relief settings consistently within +/- 3% of setpoint during successive surveillance testing, however, variances outside of'the required +/- 1% of
DUKE POWER COMPANY / CATAWBA NUCLEAR STATION-z,1 PIR 1-C90-0075/Special1 Report wPage 5 setpoint have occurred.. A Safety ' Evaluation will evaluate the effect of relief setting varying +/- 3% of setpoint due to the uncontrollable variables that affect valve performance.. These efforts will be to evaluate the' acceptability-
.of as-found (Wyle Laboratories as-received) tests varying;+/- 3% of-setpoint.
However, it' is expected that the as-lef t (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 involving an operability question. This. position, if determined acceptable, should.be documented ~in theiT/Ss; T/S Bases, or T/S Interpretations for Safety / Relief:
Valves.
The. inability;of safety valves to meet. performance of the +/- 1% test requirement is. identified as a recurring. problem.
The recommended and in-progress evaluation and the Westinghouse Owners Group efforts is expected to address these concerns and implement changes that will improve safety valve performance. A review of the Operating Experience Program (0EP) database revealed no previous events where equipment setpoints varied greater than their designed tolerance.
This report is being submitted as a Special Report in that_setpoint drift l
greater than the +/- 1% required setting range could occur in subsequent
~as-found surveillance test. The-cause of the setpoint drift could not be l
attributed to a known condition that would result in the inoperabilityLof the.
PSVs. -The setpoint drift is attributed to the effect that normal plant operation and system conditions can have on the ability of the PSV to consistently relief-at the -required s'etting. Therefore, no T/S violation has u
i occurred.
This report is being submitted as potentially 10CFR Part 21 reportable; the PSV is considered a " basis component"; a drift in setpoint greater than +/-'1% is considered a " defect"; and the' existence of a " substantial safety haza rd" is dependent upon the affect this setpoint 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
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was identified within manufacturer purchase specification CNS 1205.09-00-0001, Pressurizer and Main Steam Safety Valves.
Setpoint drifts'outside-+/- 1% of L
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.
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PLANNED 1)
A Safety Evaluation will be performed to address setpoints drifts of
+/- 3% as an acceptable condition on the PSVs.
DUKE POWER COMPANY / CATAWBA NUCLEAR STATION PIR 1-C90-0075/Special Report Page 6-4' 4
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Station Documents-(T/Ss, T/S Bases, or T/S Interpretations) will be:
revised / developed to incorporate the results of'the Safety Evaluation.
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The Safety Analysis of this report will be revised to incorporate the results of the safety evaluation.
i SAFETY ANALYSIS s
Variances ~in the PSV's setting greater than +/ 1% of setpoint.have.not been.
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considered,in the present safety' analysis assumptions. The'PSVs are required to function'to limit NC System pressure during incidents involving a' decrease in heat removal by.the, secondary. system, decrease in Reactor coolant flow rate,'and
.i anomalies in reactivity and power distribution.
A safety evaluation will be.
-j performed to address variances in PSV settings:to +/- 3% of setpoint. :The results of the safety evaluation will be incorporated into the Safety Analysis 4
of this report.
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