ML20044A328
| ML20044A328 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 06/22/1990 |
| From: | William Cahill, Mcgee G TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LER-90-016, LER-90-16, TXX-90192, NUDOCS 9006280338 | |
| Download: ML20044A328 (11) | |
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-7 7 File # 10200 910.3 I
915.6 1UELECTRIC Ref. # 50.73 50.73(a)(2)(ii) m i i
$nI,0$ r,,,ui,., June 22, 1990 U.- S. Nuclear Regulatory Commission ;
Attn: Document Control Desk l Washington, D. C. 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION DOCKET N0. 50-445 OPERATION PROHIBITED BY TECHNICAL SPECIFICATIONS 1 LICENSEE EVENT REPORT 90-016-00 l l.
Gentlemen:
. Enclosed is Licensee Event Report 90-016-00 for Comanche Peak Steam Electric l Station Unit 1, "Three of Four Steam Generator Atmospheric Relief Valves
! Inoperable Due to insufficient Stroke Length Settings."
l Sincerely, g' _
William J. Cahill, Jr.
JRW/daj '
Enclosure 1 1
c - Mr. R. D. Martin, Region IV I Resident Inspectors, CPSES (3) l l
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i 9006280338 900622 5 PDR ADOCK 0300 S . , , . 400 North Oltve Street LB81 Dallas. Texas 75201 l
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ESilMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATON 00LLECTON REOUEST: 60.0 HRS. FORWARD COMMENTS REGARDING BU@EN ESWATE T THE RE @S AN RU RTS MANAGEMENT LICENSEE EVENT REPORT (LER) BRANCH (P-630). U.S. NUCLE AR REQULATORY COMMIS$10N. WASHINGTON.
DC. 20666. AND TO THE PAPERWORK REDUCTON PROJECT (31600104).
OFFICE OF MANAGEMENT ANO BUDGET. WASHINGTON.oC.20603.
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On May 21,1990, Comanche Peak Steam Electric Station Unit 1 was conducting capacity L testing to verify that the Steam Generator Atmospheric Relief Valves (ARVs) stroked fully, using l feodwater flow increase as a qualitative indicator of valve stroke. Following the test, Engineering determined that valves 1-PV-2325,1-PV-2326 and 1-PV-2327 did not provide the minimum required steam flow capacity to support the Design Basis Accident Analysis. As a
- result, the three ARVs.were declared inoperable resulting in Unit 1 entry into Technical l Specification Limiting Condition for Operation (LCO) 3.0.3. The valves were subsequently I
calibrated and declared operable allowing exit from LCO 3.0.3.
- The event resulted from two causes: 1) The pneumatic controls for three of the four ARVs had drifted out of calibration, and 2) the specified stroke length for two of these three ARVs was
- l. reduced'due to an inadequate review and approval process for Instrumentation & Control (l&C)
L data calibration sheets Corrective actions include setting the valves to the appropriate ,
- configuration and conducting an evaluation to establish the frequency for verifying ARV stroke l length. Also, the calibration data sheets were corrected and the 1&C program for the revision of calibration data sheets had previously been revised to require the Supervisor, I&C Engineering to review any changes to design requirements.
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ESTIMATED 9URDEN PER RESPONSE TO COMPLY WITH THIS INFORWATON S
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0IO Ol2 OF 110 !i w r. - . . . i Nw r ,m 3.AA.> ii ri la DESCRIPTION OF THE REPORTABLE EVENT A. PLANT OPER ATING CONDITIONS PRIOR TO THE EVENT l l
At 1930 on May 23,1990, Comanche Peak Steam Electric Station (CPSES) Unit 1 was in Mode 3, Hot Standby, with Reactor Coolant System (Ells:(AB)) temperature and pressure at 557 degrees Fahrenheit and 2250 pounds per square inch gage, respectively, B. STATUS OF STRUCTURES. SYSTEMS. OR COMPONENTS L
THAT WERE INOPERABLE AT THE START OF THE EVENT I AND THAT CONTRIBUTED TO THE EVENT The reportable event included plant operation for approximately 64 days with three of four Steam Generator Atmospheric Relief Valves (ARVs) (Ells:(RV)(SB)) unable to = l support the Design Basis Accident Analysis. Therefore, three ARVs were inoperable i at the start of the event.
l L C. REPORTABLE EVENT CLASSIFICATION (S)
L L Any operation or condition prohibited by the plant's Technical Specifications.
Any event or condition that resulted in the nuclear power plant being in a condition -
that was outside the design basis, h . D. NARRATIVE
SUMMARY
OF THE EVENT. INCLUD;NG DATES ANb l
APPROXIMATE TIMES At 0945 on May 21,1990, CPSES Unit 1 was in Mode 1, Power Operation, operating at 35 percent reactor power. Performance and Test engineers (contractor and utility, non licensed) had _ received permission from Operations to begin a test of the ARVs.
The test was intended to verify that the ARVs stroked fully, using feedwater flow increase as a qualitative indicator of stroke, Revision 0 of the ARV Capacity Test procedure contained " Acceptance Criteria" requiring that each valve will open and close under normal hot steam conditions, in addition, the test procedure contained
" Review Criteria" requiring that each valve will have the capacity to pass steam flow equivalent to 2.5 (+1,-1) percent of the total rated steam flow. " Acceptance Criteria" provide observable results to be used to judge the acceptability (SAT or UNSAT) of I
Encio.sure to TXX 90192 NRC FORM 3664 U.S. NUCLE AR REQULATORY COMMISSON APPROWO OM8 NO. 3%0104 +
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0IO Ol3 OF 110-1..n. . - . - .o.. - - NRc o, the task while " Review Criteria" are utilized as an indicator of potential problems rather than as a limiting condition. " Review Criteria" do not necessarily render a test-unsatisfactoryif not achieved.
The test procedure was started at 0955 and completed at 1400. The measured 1 capacities for ARVs 1 PV 2325,2326,2327 and 2328 were approximately 2.8,2.4, ,
3.8, and 4.7 percent, respectively (as referenced to 100 percent steam flow, without pressure corrections). The deviation from the " Review Criteria" for the last two valves u was documented and engineering was requested to evaluate the operability impact -
of the apparent excess capacity.
While Engineering performed their review, instrument and Control (l&C) technicians (utility, non licensed) were sent to measure the stroke lengths for each valve. The design specification for the stroke length was 1-3/8 (+1/16, -0) inches. The measured stroke lengths for 1 PV 2325 through 2328 were 1-3/32,1-1/16,1-5/16, and 1-3/8 inches, respectively. The differences in ARV capacity appeared to correlate with the differences in stroke length. ,
i: The Engineering review of the test results determined that valves 1-PV-2327 and 2328 did not have excess capacity, instead, it was determined that with the L exception of 1-PV 2328, the ARVs did not appear to provide the minimum required steam flow capacity to support the Design Basis Accident Analysis for the Steam
. Generator Tube Rupture (SGTR) event. At 1930 on May 23, the three ARVs were declared inoperable resulting in Unit 1 entry into Technical Specification Limiting D
Conditic,n for Operation (LCO) 3.0.3 when the LCO for Technical Specification l 3.7.1.7 (requires a minimum of two operable ARVs) could not be met. Concurrent l with declaring the three ARVs inoperable, actions were taken to commence plant l cooldown from Mode 3 to Mode 4, Hot Shutdown, as required by Technical L Specification LCO 3.0.3. (Unit 1 had entered Mode 3 on May 22 following a planned l; reactor trip to perform a Loss of Offsite Power Test).
l&C completed the valve calibration (bench set) of 1-PV-2327 with the valve stroke length reset to 1-3/8 inches and at 2231 on May 23, Operations successfully completed the Technical Specification required surveillance test allowing 1-PV-2327 I to be declared operable. This allowed Unit 1 to exit LCO 3.0.3 at 2236 and enter the L 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Action Statement for Technical Specification 3.7.1.7 (2 ARVs operable). At
! 2055 on May 23, the Nuclear Regulatory Commission (NRC) Operations C. .ter was
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r notified of the inoperabie ARVs via the Emergency Notification System in accordance- ,
with 10CFR50.72(b)(2)(l). A followup call to the NRC was subsequently made at-2241 which identified that LCO 3.0.3 had been exited.
The valve calibration (bench set) and Technical Specification required surveillance testing for 1-PV-2326 and 1-PV-2325 were subsequently completed and the valves i declared operable at 0510 on May 24 and at 1919 on May 25, respectively. The Action Statement for Technical Specification 3.7.1.7 was exited and a Tracking-Limiting Condition of Operation Action Requirement was initiated to ensure the ARV ,
Capacity Test was performed when appropriate operating conditions were achieved.
After resetting the ARV's stroke length to 13/8 inch and successfully completing the Technical Specification required surveillance testing of the three ARVs, the l Operations Manager (utility, licensed) gave permission for Unit 1 entry into Mode 2, !
Startup, and Mode 1.
The ARV Capacity Test Procedure " Review Criteria" was changed to " Acceptance
' Criteria", The new " Acceptance Criteria" for ARV capacity was changed to require i 779,000 to 968,000 pound mass per hour (LBMH) On May 26, once necessary plant conditions were achieved, a second ARV Capacity Test was begun; however, it-was only partially completed. The data was inconclusive as the test was interrupted by a reactor trip caused by a feedwater control valve (Ells:(FCV)(SJ)) problem-unrelated to the test.
On May 29, the ARV Capacity Test was reperformed. Measured flow rates were
. 779,000; 585,000; 7_10,000; and 668,000 LBMH, Based on an Engineerirg analysis, the ARVs were determined to be acceptable for the existing plant condiilons (50 t percent power or less).
p During the period from May 30 to June 5, Unit 1 remained at 50 percent power while assessing operations (as committed to the NRC prior to increasing power to 75 percent). I&C checked the stroke lengths following the May 29 test performance and found the stroke of 1-PV-2326 (flowrate of 558,000 LBMH) to be less than specified (1-7/32 versus 1-3/8 inches) when operated from the control room. Previously, I&C q had stroked the ARVs locally using instrument air when setting the mechanical travel ,
stops at 13/8 inches. The control loop was recalibrated and the valve stroked to the !
correct length from the control room. All ARVs were verified to stroke the correct length from the control room.- The valve vendor (Fisher Controls) was also brought on site to assist in valve calibration.
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0IOsoI5 OF 110 to -. . ,% o .umo o,. 3 A .) o r) i L On June 7, Engineering re established the design basis for the ARVs. The ARVs are required to remove a minimum of 756,000 LBMH at 1200 pounds per square inch-atmosphere (PSIA) to ensure that the capacity is sufficient to cool down the plant j t during a loss of off-site power such that at 100 PSIA the valves pass a minimum of ;
L 62,150 LBMH, Thisvalue (62,150 LBMH) bounds the capacity required for a SGTR event, such that there will be a sufficient cooldown rate to depressurize the reactor c coolant system, allowing an operator to stop the leakage prior to the affected steam 1 generator filling with liquid. In addition, the maximum relief capacity of each ARV is i l; 968,400 LBMH so as not to exceed the maximum cooldown rate specified by the l steam supply vendor. The bounding analysis for this maximum value is the stuck l l: open main steam safety valve, per FSAR Chapter 15.1.5. The maximum value is also used in the radiological release calculations for determining the off-site dose i consequences following an SGTR event.
Engineering obtained confirmatory information from the valve vendor stating that !
steam flow through the valves can be correlated to valve stroke within 3 percent.
Thus, valve stroke length is the appropriate parameter to measure in determining ;
p valve operability, rather than using a secondary heat balance (due to the difficulty in l l accounting for the flows in all available steam flow paths). Engineering issued a ;
design modification to increase valve stroke an additional 1/16 inch to 17/16 (+1/16,
-0) inches.- This was done to ensure that margin exists between the minimum and maximum capacity requirements and the actual valve setting. A comparison of steam flows calculated at the revised stroke lengths to the minimum and maximum required flow rates bounds the 3 percent tolerance with margin remaining. The design modification was implemented and the post modification testing (verification of stroke length) and Technical Specification required surveillance testing were successfully ,
completed. The ARVs were then declared fully operable (to 100 percent power) on
' June 14,.1990, i E. THE METHOD OF DISCOVERY OF EACH COMPONENT OR SYSTEM FAILURE OR PROCEDURAL OR PERSONNEL ERROR The inoperability of the three ARVs was discovered by the Engineering review of the test results for the ARV Capacity Test.
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Encio,sure to TXX 90192 NRC FORM 3664 U.S. NUCLE AR REOULATORY COMMIS$ON APPROVED OM8 NO 31640104 3-ESTIMATED BORDEN PER RES ST COMPLY WITH THIS INFORMATON LICENSEE EVENT REPORT (LER) @n'ty" " 1T*,'; 1%'"ffo',7,'"o " 7 ,' f f g 1
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- 11. COMPONENT OR SYSTEM FAILURES A. FAILURE MODE. MECHANISM AND EFFECT OF EACH FAILED COMPONENT !
Three of the four ARVs were determined to be inoperable since the valves did not stroke sufficiently to support the Design Basis Accident Analysis. This resulted from the stroke length specified on the l&C data calibration sheets for two of the ARVs (1-PV 2325 and 1 PV 2326) being lower than the established design requirements combined with the fact that the pneumatic controls for 1 PV 2325,1-PV 2326 and ;
j 1-PV-2327 had drifted out of calibration, j 1
L B. CAUSE OF EACH COMPONENT OR SYSTEM FAILURE i i
l The causes for the inoperable ARVs are described in Section IV.
C. SYSTEMS OR SECONDARY FUNCTIONS.THAT WERE AFFECTED BY FAILURE OF COMPONENTS WITH MULYlPLE FUNCTIONS Not applicable - Although the ARVs are containment isolation valves and provide a l containment isolation function, the abil'ty to perform this function was not affected by l the problems identified in this event.
l.
L D. COMPONENT INFORMATION L ARVs 1-PV-2325,2326, and 2327 Manufacturer: Fisher Controls Model Number: 667-EWP l 1
Ill. ANALYSIS OF THE EVEN T A. SAFETY SYSTEM RESPONSES THAT OCCURRED i Not applicable - There were no manual or automatic safety system responses as a result of this event.
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Enclosure to TXX 90192
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B. DURATION OF SAFETY SYSTEM TRAIN .NOPERABILITY The ARVs do not directly correlate with ihe term " safety system train." However, the ARVs do provide a safety function and were required to be operable in accordance with Technical Specifications beginning on March 20,1990 when Mode 3 was .,
entered (approximately 64 days passed until the inoperability was identified by ARV '
Capacity Testing). Upon discovery of the inoperability at 1930 on May 23,1 PV-2327,2326,2325 were restored to an operable status within 3,10, and 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> -
. (approximate times), respectively.
C. SAFETY CONSEQUENCES AND IMPLICATIONS OF THE EVENT s
A minimum ARV capacity is required for the following reasons:
L 1) Sufficient ARV capacity is needed to assure a timely plant cooldown for the L mitigation of the' design basis SGTR accident. This cooldown is required in
'~
order to allow the primary to secondary leakage to be terminated by an operator 6 prior to the affected steam generator (Ells:(SG)(AB)) filling with liquid.
- 2) Sufficient ARV capacity is required to allow for cooling the plant from Hot Standby (Mode _3) to Residual Heat Removal (RHR) (Ells:(BP)) system cut in .
conditions in the event of a loss of offsite power. The cooldown must be accomplished prior to the depletion of Condensate Storage Tank (CST)
(Ells:(TK)(KA)) inventory.
An analysis was performed of the "as found" capacities during the initial performance of ARV Capacity Test and the results are summarized below:
In the absence of a single active failure, the reduced ARV capacity would not 1 1) have prevented the ARVs from performing their intended safety function during the mitigation of a SGTR accident occurring while the plant was operating at or below 100 percent power. In addition, other systems, such as the radiation imnitors (Ells:(MON)(IL)) (providing for early detection of an SGTR) and the steam dumps (Ells:(RV)(SB)) (providing an alternate heat removal path), were available for use in the mitigation of an SGTR event.
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' Encio,sure to TXX 90192 -
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- 2) In the absence of a single active failure, the reduced ARV capacity would have been sufficient to allow the ARVs to perform their function of cooling the Reactor Coolant System to RHR system cut-in conditions prior to depleting the CST following a plant trip from power levels at or below 100 percent.
, IV.- CAUSE OF THE EVENT .
L
.The stroke length of three of the ARVs was found to be lower than design requirerttants .
resulting in their inoperability. This occurred due to the following two causes:
ROOT CAUSE-1 ,
?
l The pneumatic controls for ARVs 1-PV-2325,1 PV 2326 and 1 PV-2327 had drifted out of calibration.
ROOT CAUSE - 2 l' Prior to October 1989, an inadequate review and approval process existed for I&C '
calibration data sheets. This process allowed a revision to the " stand alone" data sheets for 1 PV-2325 and 1-PV-2326 which reduced the specified stroke length from L- 1-3/8 inch to 1-1/4 inch, 1
CONTRIBUTING FACTOR The analytical determination of the 1-3/8 inch stroke length (verified by the valve vendor) 1 li resulted in a steam flow very near the minimum value required by the accident analysis E
Based on the uncertainties associated with correlating valve flow capacity and valve stroke length, it would have been prudent to establish a valve stroke setting with adequate l tolerances.
l l.
V. CORRECTIVE ACTIONS ,
A. CORRECTIVE ACTIONS TO PREVENT RECURRENCE I
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ROOT CAUSE - 1 L
The pneumatic controls for ARVs 1 PV 2325,1 PV 2326 and 1-PV 2327 had drifted out of calibration.
CORRECTIVE ACTION i An evaluation will be performed to establish the frequency for verifying ARV stroke 1 length to ensure that the valves are within tolerance. Valve stroke measurement will be performed after opening the valve from the control room via the Manual / Auto l l station.-
l:
L ROOT CAUSE - 2
- l' Prior to October 1989, an inadequate review and approval process existed for i calibration data sheets, a
CORRECTIVE ACTION The l&C program for revision of calibration data sheets was revised in October 1989 to require the Supervisor, I&C Engineering (in lieu of a Shop Supervisor) to review L
any chanr:es to design requirements. This program change is adequate for control of future revisions of calibration data sheets.
CONTRIBUTING FACTOR L The analytical determination of 13/8 Inch stroke length resulted in a steam flow rate j very near the minimum value required by the accident analysis. ,
L E
CORRECTIVE ACTION The appropriate calculation has been revised and a design modification has been l L~ implemented to increase valve stroke an additional 1/16 inch. This was done to l ensure margin exists between the minimum and maximum capacity requirements l .and the actual valve settings. ;
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B. CORRECTIVE ACTION TAKEN ON GENERIC CONCERNS IDENTIFIED AS A DIRECT RESULT OF THE EVENT GENERIC IMPLICATION Since a backfit review was not required as part of the I&C review and approval process change in October 1989, additional deficiencies may exist for " stand alone" l data calibration forms.
CORRECTIVE ACTION To provide additional assurance that similar deficiencies do not exist, a representative sample of quality related " stand alone" calibration data sheets was
[- . reviewed to assess consistency with design requirements. Unlike the " stand alone" ,
data calibration sheets which provide the sole source of project documentation for the calibration data, data calibration sheets with supporting procedures, documentation, etc. are considered validated and there is no need to include them in the sample, No deficiencies were found during this evaluation.
VI. PREVIOUS SIMILAR EVENTS There have been no previous similar events reported pursuant to 10CFR50.73.
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-Vll. ADDITIONAL INFORMATION All times are approximate and Central Daylight Savings Time (CDT).
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