05000482/LER-1998-001, :on 980304,noted That Pressurizer Code Safety Valves Indicate Five Random Failures in Last Twelve Tests. Caused by Incapability of Valves to Meet Restrictive Tolerance of +/-1% of Valve Pressure.Ts Will Be Amended
| ML20238E755 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 08/27/1998 |
| From: | Angus M WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | |
| Shared Package | |
| ML20238E753 | List:
|
| References | |
| LER-98-001, LER-98-1, NUDOCS 9809020165 | |
| Download: ML20238E755 (9) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(viii)(A) |
| 4821998001R00 - NRC Website | |
text
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'NRC FORM 366 U.S. Nt! CLEAR REGURATO7.Y COMMISSION APPROVED BY OMB NO.3150-0104
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WOLF CREEK GENERATING STATION 05000482 l
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Pressurizer Code Safety Valves Testing Outside of Technical Specification Allowances EVENT DATE (5)
LER NUMBER (b)
REPORT DATE (7)
OTHER FACILITIES INVOLVED (8)
MONTH DAY YEAR YEAR SEQUEN14AL REV MONM DAY YEAR f ACluTY NAME DOCKET NUMBER NUMBER NUMBER 03 04 98 98 001 01 08 27 98 F ACluTY NAME DOCKET NUMBER UtthAllNG THIS kEFukI IS SVbM11TLD FURSUANT IU 1HL RLyUlkLMLNIS Ok 10 Ctk bl (check ODS Or motel (11)
NODE (9)
MODE 1 20 402(b) 20 40d(c) 50 73(a)(2)0v) 73 71(b)
FUWLk 20 405(a)(1)0) 50.36(c)(1) 50 73(a)(2)(v) 73 71(c) 100 20.405(a)(1)(n) 50.36(c)(2) 50 73(a)(2)(vn)
X OTHER LEVEL (10) percent 20.405(a)(1)0ii) 50 73(a)(2)(i) 50.73(a)(2)(viii)(A)
Voluntary l
20.405(s)(1)0v) 50 73(a)(2)(n) 50 73(a)(2Hvm)(B)
,,,;~
20 405(a)(1)(v) 50 73(a)(2)09) 50 73(a)(2)(x)
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Michael J. Angus Manager, Licensing and Corrective Action 316-364-8831~ Extension-4077 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
cAUSE
SV6 TEM COMPONENT MANUF ACTURER RE 1 d CAut4 SV6 TEM cQMPQNENT MANUFACTURER HE T d j
X AB AB-V Crosby YES k
l 1
SUPPLEMENTAL REPORT EXPECTED (14)
MONTH DAY YEAR l
EXPECTED YES X
NO SDsMISSION DATE (15) m1wi nes s Wolf Creek Nuclear Operating Corporation's (WCNOC) Technical Specification 3.4.2.2 states, "All pressurizer code' safety valves shall be operable with a lift setting of 2485 psig +/-
]
1%."
Surveillance test results of Wolf Creek Generating Station's (WCGS) pressurizer code j
safety valves indicate five random failures in the last twelve tests over a time period of i
four refueling outages.
The most recent failure was from a valve removed in Refueling Outage (RF) IX, Fall 1997, and tested in February 1998.
WCNOC reviewed the pressurizer code safety valve history and concluded that the root cause is that the valves are not capable of consistently meeting a restrictive tolerance of +/-
l% of valve test pressure with currently available industry test methodology and equipment.
This conclusion is consistent with WCNOC test data, NRC Information Notice 91-74 findings, Westinghouse root cause conclusions, industry experience, and history of WCNOC main' steam safety valve performance. The corrective action is to pursue a change to i
the Technical Specification set point and set point tolerance for the pressurizer safety
- valves, j
9309020165 980827 PDR ADOCK 05000482l S
.U.S. MUCLEAR REGULATORY COMMISSION (6-99)
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TEXT CONTINUATION l'ACILITY NAME (1)
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YEAR SEQUENTIAL REVISION NUMBER man Wolf Creek Generating Station 05000482 98 001 01 2
OF 8
YL%T (lfmore space is requsred, use additionalcopies of bRC f arm 3664 (E7)_
Pl nt Conditions Prior to the Event:
MODE = 1 Reactor Coolant Pressure: = or < 2235 psig Reactor Power: 100 percent Ba-is for Deportability:
Wolf Creek Nuclear Operating Corporation's (WCNOC) Technical Specification 3.4.2.2 states, "All pressurizer code safety valves [AB-V] shall be operable with a lift setting of 2485 psig
+/-l%."
Technical Specification Surveillance Requirement 4.4.2.2,
- states, "No additional requirements other than those required by Technical Specification 4.0.5."-
Technical Specification 4.0.5 requires testing in accordance with ASME OM-1987 Part 1 (ASME OM-1).
WCNOC supplements the ASME requirements with guidance from NUREG-1482,
" Guidelines for Inservice Testing at Nuclear Power Plants" and ASME Code interpretations.
Test results of WCNOC's pressurizer code safety valves indicate five random as-found lift test failures in the last twelve tests over a time period of four refueling outages.
The most recent test failure was from a valve removed in Refueling Outage (RF) IX, Fall 1997, and tested in February 1998.
The historical test data provides evidence of random as-found set point test failures.
Partial disassembly and inspection performed during testing cycles showed that the valves were in good condition and did not find any apparent explanation for deviation between as-left and as-found settings.
It is, therefore, impossible to conclusively determine the time of degradation for any of the valves that failed the as-found set point lift tests.
NUREG-1022, " Event Reporting Guidelines 10 CFR 50.72 and 50.73," Revision 1 states in part that, "... discrepancies found in Technical Specification surveillance tests should be assumed to occur at the time of the test unless there is firm evidence, based on-a review of relevant information (e.g.
the equipment history and the cause of_the failure) to believe that the discrepancy occurred earlier."
Therefore, it is assumed that the degradation occurred at.the time of the tests and thus WCNOC is reporting this event on a voluntary basis.
De*cription of Event:
Test results of WCNOC pressurizer code safety valves indicate five random failures in the last twelve tests over a time period of four refueling outages.
NRC Information Notice 91-74 discusses pressurizer safety valve set points.
An excerpt from the document follows:
"The industry has continued to find that the set points of PSVs and main steam safety and relief valves can be unreliable.
Recently, the NRC staf f reviewed data from the Nuclear Plant Reliability Data System (NPRDS) and found that, historically, as-found PSVs in the U.S.
plants have failed over 40 percent of the set point tests performed.
The NRC believes that the number may actually be higher because of unreported fallures."
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.U.S. NUCLEAR REGULATORY COMMISSION (6-98)
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TEXT CONTINUATION 4
FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
FAGE 13)
YEAR SEQUENTIAL REVI510N N'JMBER M DER Wolf Creek Generating Station 05000482 98 001 01 3
OF 8
ILXi (if more space os requsred, use addatonal copses of hkC iorm !6M) (87}
The data.at WCNOC for the last four post refueling outage tests is consistent with what is occurring in the industry.
That is, there' is an ~40% failure rate on set point testing.
But during the last.three post refueling outage testing only two failures in nine tests (~22%) occurred, which is a significant improvement, and less than the 40%
industry average in 1991.
Also, during the last three post refueling as-found set point tests the set pressure had deviated less than 2% from the nominal set pressure. -This is also an improvement over the post RF VI as-found set pressures.
The last four refueling outage test results are discussed below.
Valve test failures are listed below:
Refuel Outage Valve Serial No.*
AF to Nominal Delta
(% of Nominal)
REVI BB8010A 0001
- - 2.13 RFVI BB8010B 0039
+1.40 RFVI BB8010C 0003
- - 2.70 RFVIII-BB8010B 0039
+1.97 RFIX BB8010A 0002
- - 1.81
- These are the last four digits (XXXX) of the serial number: N60446-00-XXXX RF VI In 1991, WCNOC changed from testing on nitrogen to testing on steam.
So, 1993 was the
- first time the as-found tests were conducted af ter the valves were as-lef t on steam.
The
. set pressure test results from posL RF VI testing are questionable.
Some differences that may be attributed to the failures are:
o. In 1991 the actual ambient test temperatures ranged from 89 'F to 103 'F; e
the valve body test temperature range was 305 'F to 333 *F, except for the last test on valve N60446-00-0039, where the temperature was 199 'F; and three-informational lifts were performed with nitrogen on each valve after jack-and-e lap and seat leakage tests were completed post RF IV. This likely changed the as-left set pressure actual condition from the post RFIV recorded as-left lift test pressure.
RF VII As-found post RF VII valve tests were in tolerance.
RF VIII Both as-found set pressure tests for valve SN-0039 for post RF VI and RF VIII testing were found to be higher than the +1% Technical Specification tolerance, and are considered failures.
In addition, the as-left to as-found delta % has been high, 1.16 and 2.47 for post RF VI and RF VIII testing, respectively.
To adequately address the performance of
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lU.S. NUCLEAR REGULATORY COMMISiloN l
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TEXT CONTINUATION l
FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
FAGE (3)
YEAR SEQUENTIAi, FIVISION NUMBER NUMBEA Wolf Creek Generating Station 05000482 98 001 01 4
OF 8
ILXT (if more space os requtred, use addasonalcopies of A RC form 36M) (E7) l valve SN-0039, the data. from post RF IV (7/91) to post RF VIII (7/96) testing was reviewed.
I i
As mentioned in the post RF VI test discussion, the nitrogen lifts after the post RF IV l
testing likely had some impact on the as-found set pressure testing performed post RF VI.
l-However, the difference between the aborted test pressure and final (4th) lift of 47 psig
(-2%) appears more significant.
The four tests were performed' without adjustments to the valve. Testing performed post RF IV and subsequent testing of the valve performed post RF VIII indicate that the valve performance is more repeatable than exhibited in post RF VI testing. The valve lifted within 14 psi on the two lifts performed in post RF IV testing.
Although above the allowed +1% tolerance, the as-found test and the two consecutive lifts
(
thereafter performed without adjustment had a maximum range of 11 psig for post RF VIII testing (2534 psig, 2523 psig, and'2529 psig, respectively)
Af ter this series of tests the valve was disassembled and inspected and, per the Certificate of Conformance, no problems were found.
The valve was reassembled, reset and tested. Two more lift tests were performed in the post RF VIII testing to establish an as-left pressure which resulted in a range of only 8 psig (2496 psig and 2488 psig).
The reliability of the _ valve demonstrated by post RF IV and post RF VIII testing leads one to question the post RF VI l
test results.
It is noted that during the post RF VI testing, the valve initially started leaking at 2400 psig.
The pressure was increased at a ramp rate of 2 psi /see until the j
test was aborted at 2520 psig because of excessive leakage.
The ramp rate was increased to 9 psi /sec for test 2 and the valve lifted at 2512 psig.
The ramp rate was decreased to l
l 5 psi /sec for test 3 and the valve lifted at 2479 psig.
The ramp rate was again decreased I
to 3 psi /sec and the 4th and final lift occurred at 2473 psig.
Tne third and fourth tests I
of this sequence, within 6 psig, are more in line with the as-left 2491 psig from post RF l
IV testing.
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RF IX During post RF IX testing for valve SN-002 the as-found to nominal failed low (-1.81%) and i
the previous as-left to as-found was low (-1.97%).
1 After failing the first as-found post RF IX lift test at 2440 psig, the valve was tested two more times without any adjustments.
The pressures at which the valve opened were acceptable at 2476 and 2464 psig.
These values are higher than the first as-found test and closer to the previous as-left pressure of 2489 psig. After an adjustment was made to the valve to increase the set pressure, the valve opened with repeatability at pressures of 2499 and 2487 psig.
During the jack and lap process to assure seat tightness, the valve nozzle seat step was found to be at the minimum value.
The seat step was re-l established and the valve re-tested.
The valve performance was again repeatable at 2474 and 2467 psig.
Since the valve performance was demonstrated to be repeatable after the low as-found pressure 'of 2440 psig, it is believed that this result was an anomaly and is unexplainable.
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- IvlsION NUMBER m EER Wolf Creek Generating Station 05000482.
98 001 01 5
TL%1 (ymore space as requsred, use additw,wl copies of ARCIurm 566A) (17)
Root Cause:
Pressurizer Safety Valve reliability to meet the as-found set point Technical Specification tolerance of +/- 1% of set pressure has improved since Refuel VII, when we began using steam to conduct as-found lift tests and the valves were being tested to ASME/ ANSI OM-1 requirements.
However, in spite of improved test controls, it continues to be a challenge to meet acceptable as found pressures and satisfy the requirement to demonstrate as-left repeatability within the Technical Specification tolerance with two consecutive lift tests.
Test records indicate that the valves performance have been repeatable within a tolerance of +/- 2% for the last three post refueling outage as-found tests.
The root cause is that the valves are not capable of consistently meeting a restrictive tolerance of 4/-
1% of valve test pressure with currently available industry test j
methodology and equipment. This conclusion is consistent with WCNOC test data, NRC IN 91-74 findings, Westinghouse root cause conclusions, industry experience, and history of WCNOC Main Steam Safety Valve performance.
Corrective Actions
The as-found set point lift test failure was documented in Action Request 27717 Performance Improvement Request (PIR) 98-0743 and Deportability Evaluation Request (RER)98-010 were initiated to evaluate root cause, corzective actions, and potential deportability.
The set pressure for safety valve SN N60446-00-0002 was adjusted and the valve retested to certify that as-left set pressure was within Technical Specification tolerance.
A subject matter expert from Westinghouse was contacted and assisted in the review of the equipment history and root cause evaluation.
The review of historical valve test documentation revealed that no non-conformance document had been initiated for one PSV test failure in 1996.
PIR 98-0758 and RER 98-013 were ir4itiated to further review the 1996 failure and correct the root cause and any programmatic weaknesses.
The long term corrective action is to pursue a change to the Technical Specification set point tolerance for the Pressurizer Safety Valves. WCNOC Engineering is currently working on the safety analysis to support revision of Technical Specification 3.4.2.2 to lower the pressurizer safety valve set point 1% and change the tolerance to +/- 2%.
The safety I
analysis is expected to be complete by November 20, 1998.
S$fety Significance:
June 1993 Evaluation (Refuel VI)
The as-found opening pressures of the PSVs in June 1993 were as follows:
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05000482 98 001 01 6
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i L% I ({fmore space ss reqwrrd, use oddinonalcopses of ARCiorm 3664 (l7)
Valve Serial As-Found Percent Difference Number Opening-Pressure from Set point (2485 psig)
N60446-00-0001 2432 psig
- - 2.13 N60446-00-0039 2520 psig
+1.40
)
N60446-00-0003 2418 psig
- - 2.70 1
The as-found opening pressure of each individual PSV was less than the as-round value in July 1996.
Therefore, with respect to over-pressurization, the June 1993 conditions are clearly bounded by the July 1996 conditions.
l
.The as-found conditions in June 1993 exceeded the minimum allowed opening pressure.
PSVs 001 and 003 opened at -2.13% and -2.~10%,
respectively.
The analysis of the Inadvertent Operation of the ECCS During Power Operation and Inadvertent Opening of a Pressurizer Code Safety Valve are potentially impacted by the PSVs opening below the minimum allowed pressure.
Inadvertent Operation of the ECCS During Power Operation I
1 This event addresses the consequences of spurious actuation of the safety injection (SI) t
- - tem.
Following actuation of the SI system, both trains inject borated water into the sld legs of the Reactor Coolant System (RCS).
The initiating event, a spurious SI signal, would also be expected to generate a reactor trip, although an immediate reactor trip is not credited in the analysis.
The analysis is performed to bound the transient i
conditions with or without an immediate reactor trip.
Depending on the control system in operation, core power and RCS temperature remain near the initial nominal condition or decrease during the event, and RCS flow remains constant.
A decrease in RCS pressure is the only condition that may occur which would adversely affect departure from nucleate boiling (DNB).
However, for the decrease in RCS pressure which may occur, the effects are generally more than offset by beneficial changes in power and temperature. The net effect j
is a DNB ratio (DNBR) that remains near the initial value or increases through the event.
Hence, tne safety analysis DNBR limit is not challenged during the analyzed transient.
The main concern that results from an inadvertent ECCS actuation event is that associated with pressurizer overfill.
The pressurizer water volume increases for this event as a result of the SI flow.
Ultimately, operator action is required to terminate SI flow, and thereby, terminate the water in-surge to the pressurizer.
For this event, it is conservative with respect to the minimum time before reaching a pressurizer water solid condition to model conditions which minimize RCS pressure.
Minimizing the RCS pressure maximizes the SI flow injected into the RCS, due to pump performance characteristics, and thereby maximizes the water in-surge into the pressurizer.
The analysis conservatively assumes automatic pressure control systems function (i.e.,
power-operated relief valves (PORVs) and spray).
As a result, the calculated pressure in the limiting pressurizer over-fill case does not significantly exceed the PORV opening pressure of 2335 psig.
The pressure does not reach the minimum opening pressure of the pressurizer safety valves even I
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TL%T (if more space os requored. use addawnal cojnes of ARC i orm 366A) (I Tp considering the low as-found opening pressures from the June 1993 surveillance tests.
Therefore, the USAR analysis bounds the as-found conditions.
Inadvertent Opening of a Pressurizer Code Safety Valve While low safety valve set pressures are not specifically discussed, a safety valve with a set pressure that is as low as 2235 psig may be compared to an inadvertent opening.
The inadvertent opening occurs at RCS operating pressure of 2235 psig, Normal operating pressure.is 250 psig below RCS design pressure and safety valve set pressure (2485 psig)
The as-found set pressure was still 183 psig, or approximately 8.2% above RCS normal operating pressure of 2235 psig.
July 1996 Evaluation (Refuel VIII) l The Wolf Creek safety analyses model the opening pressure of the pressurizer safety valves I
at 102% of the nominal set pressure (2485 psig).
The increase of +2% above the nominal set pressure supports a 1% tolerance, as required by the Technical Specifications, and accounts for a potential 1% set point shift due to the presence of a water loop seal.
The as-found condition of PSV serial number N60446-00-0039 in July 1996, which was 1.97% above the nominal set point, was not within the 1% tolerance assumed in the analyses.
Although a 1.97% variation between the nominal set pressure and the opening pressure was not explicitly bounded by the USAR Chapter 15 safety analyses, the following provides reasonable assurance that the analyses remain conservative with respect to the combined as-found conditions of the three PSVs.
The as-found opening pressures of the PSVs in July 1996 were as follows:
Valve Serial As-Found Difference from Set Number Opening Pressure point (2485 psig)
N60446-00-001 2491 psig
+0.24 N60446-00-039 2534 psig
+1.97 N60446-00-003 2483 psig
- - 0.08 The limiting over-pressure transient analyzed in Chapter 15 of the Wolf Creek USAR is the Turbine Trip event presented in Section 15.2.3.
The applicable analysis during Cycle 8 (cycle just prior to RF.VIII) is documented in SA-92-079 Revision O.
The analysis assumed a +1% PSV uncertainty, in addition to the 41% set point shift due to water loop seal (+2%
total).
The. analysis also assumed a water loop seal purge time of 1.156 seconds, further delaying relief through the PSVs.
In order to bound the condition where all PSVs might open at the extreme condition allowed by the Tech. Spec., the PSVs are each assumed to open at the maximum value.
The peak RCS pressure calculated was 2737.5 psia compared to a 1
limit of 2748.5 psia.
Since the as-found condition of PSV 0001 and 0003 were well below the +1% uncertainty, these valves would have provided relief prior to the time they were credited in the USAR analysis.
Were these valves to open corresponding to their respective as-found conditions, the pressurization rate in the RCS would have been slowed significantly such l
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NRc TORM 366A U.S. WUCLEAR REGULATORY COMMISSION
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1 L%T (ymore space is required, use adJtitortal copies of ARC form 3664) (L1) that the final PSV (0039) would also open prior to exceeding the over-pressure limit.
Once all-- three PSVs have opened, the RCS pressure would quickly peak and ' begin to decrease.
The resulting transient is not expected to be significantly more limiting than the USAR analysis, which showed 11 psi margin to the over-pressure limit.
February 1998 (Refuel IX)
The as-found opening pressures of the PSVs in February 1998 Valve Serial As-Found Difference from Number Opening Pressure Setpoint (2485 psig)
N60446-00-0002 2440 psig
- - 1.81 N60446-00-0040 2505 psig
+0.80 N60446-00-0041 2504 psig
+0.76 The first lift for PSV N60446-00-0002 occurred at 1.8% under set pressure, This lift exceeded Technical Specification lift tolerance of 2485 psig +/- 1% by 0.8%.
The low lift pressure will not affect the ability of the valve to open and reach full rated flow as designed, and when combined with other pressurizer safeties, will accomplish its design function of maintaining RCS pressure within 110% of design.
The WCNOC accident analysis describes two events which could be impacted by pressurizer code safety valve low lift pressure:
- 1) inadvertent opening of a pressurizer code safety valve and 2) inadvertent operation of the Emergency Core Cooling System (ECCS) which are both covered in the June
{
1993 evaluation.
Conclusions for Safety Significance Section With respect to over-pressurization, the RCS pressure limit (110% of design) would be expected to be met.
With respect to minimum pressure, the Inadvertent ECCS Operation pressurizer over-fill analysis was not impacted by the as-found conditions.
Therefore, the as-found conditions of these valves would not have prevented fulfillment of the safety function, represented an unanalyzed condition, or placed the plant outside of the safety analysis design basis.
Other Previous Ocevrrences:
Three pressurizer ode safety valves failed to meet the Technical Specification surveillance requirement of +/- 1% in 1993 following RF VI.
This occurrence is discussed in this LER.
One pressurizer code safety valve failed to meet the Technical Specification surveillance requirement of +/- 1% in 1996 following RF VIII. This occurrence is discussed in this LER.
I
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Anachm:nt to WO 98-0080 Page1ofI LIST OF, COMMITMENTS i
The following table identifies those actions committed to by Wolf Creek Nuclear Operating Corporation (WCNOC) in this document. Any other statements in this submittal are provided for information purposes and are not considered to be J
commitments
Please notify the. Manager Licensing & Corrective Action at Wolf Creek
^
Nuclear Generating Station of any questions regarding this document or any associated
commitments
COMMITMENT
Date/ Event The long term corrective action is to pursue a change to the March 31,1999 Technical Specification set point tolerance for the Pressurizer Safety Valves. WCNOC Engineering is currently working on the safety analysis to support revision of Technical Specification 3.4.2.2 to lower the pressurizer safety valve set peint 1% and change the tolerance to +/- 2%.
The safety analysis i; expected to be complete by November 20, November 20,
- 1998, 1998 l
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