ML20046C082

LER 93-012-00:on 930510,reported Unusual Noise Due to Small Frictional Force Release in RCS During heat-up.Developed & Implemented Enhanced Thermal Expansion Monitoring Program. W/930803 Ltr

ML20046C082
Person / Time
Site:	Wolf Creek
Issue date:	08/03/1993
From:	Maynard O, Moles K WOLF CREEK NUCLEAR OPERATING CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-93-012, LER-93-12, WO-93-0129, WO-93-129, NUDOCS 9308090184
Download: ML20046C082 (17)

Text

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WOLF CREEK ' NUCLEAR OPERATING CORPORATION .

C Otto L. Maynard V,w F, u t riant Ormei.ve August-3, 1993 J WO 93-0129 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1-137 Washington, D. C. 20555

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Subject:

Docket No. 50-482: Licensee Event Report 93-012-00' Gentlemen:

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The attached Licensee Event - Report (LER) is being submitted as.a voluntary >

report _regarding the Noise Event experienced during heatup following the Sixth Refueling Outage. .

-1 I

Very truly yours,

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y / f ff l Otto L. aynard Vice President Plant Operations OLM/jra  !

1 Attachment l

'I cc: W. D. Johnson (NRC), w/a j J. L. Milhoan (NRC), w/a i G. A. Pick (NRC) , ' w/a -l W. D. Reckley (NRC), w/a '!

090019 ,

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, 40. Box 411/ Burkngton, KS 66839 / Phone. (316) 364-8831 P

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a 9309090184 930803 .

' An Equal Oppodunity Employer M,T/HC/ VET ' l- 1 PDh ADOCK 05000482 his ,

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NRC'FORQ 366 U.S. NUCLEAR CEGULATORY COMMIS$10N APPCOVED BY 0"B No. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

LICENSEE EVENT REPORT (LER)

FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO 6 INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, (See reverse for required ruber of digits / characters for each block)

WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150 0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) Wolf Creek Generating Station DOCKET NUMBER (2) lPAGE(3) 05000482 I 1 OF 16 TITLE (4) Small Frictional Force Release In the Reactor Coolant System Support System During Heat-up Causing Unusual Noise FVFNT DATF f5) IFR WUNRFR (6) RFPORT DATF (7) OTHFR FAEflfTfFE TWUdtVFD fRi DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACILITY NAME DOCKET NUMBER GONTH 05000 05 10 93 93 -- 012 --

00 08 03 93 FACILIM NAME CKET NUMBER 05000 OPERATING 3 TMit RFPf'RT ft tumMf7TFn PURtutWT Yn THF RFOUfRFMFWTC nF in EFE O (Ehed nne er w ei (114 20.402(b) 20.405(c) SI).73(a)(2)(iv) 73.71(b)

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POWER 0 20.405(a)(1)(li) 50.36(c)(2) 50.73(a)(2)(vii) X OTHER 2n ansta3r13rliti sn 71rnir73rti %n_71rmir?irvi11ifA1 Voluntary 20.405(a)(1)(iv) 50.73(a)(2)(ii) 50.73(a)(2)(viii)(B) 20.405(a)(1)(v) 50.73(a)(2)(iii) 50.73(a)(2)(x)

IfrFWKFF FOWTAET Fn# THft iFR (121 NAME TELEPHONE NUMBER (Include Area Code)

Mr. Kevin J. Moles - Manager Regulatory Services (316) 364-8831 ,

EnMP!rTF OWF tIWF Fn# FArH cnMPop FMT F AfI t tRF f'FtFR f RFB fM T6fK RFPORT f111 MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT s TO NPRDS - TO NPRDS bbh M&n -

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EXPECTED MnWTw cAv VFAR tHPPtFMFWTAt RFPn#T FYPFETFD (141 YES X NO (1f yes, complete EXPECTED SUBMISSION DATE).

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

On May 10, 1993, at 1507 CDT, with the unit in MODE 3, " HOT STANDBY," at approximately 540 OF the Control Room received an alam on Annunciator 98E,

" Seismic Recorder On." Loose Parts and Vibration Monitoring System (LPVMS) annunciator alams were also received. Personnel inside the Containment Building reported hearing a single loud metallic slap (sound). Personnel in the Auxiliary Building reported hearing a rumble (thunder like) sound. A Reactor Coc,lant System (RCS) leak rate test was performed and an extensive walk down of the RCS, connecting systems, and the Containment, Auxiliary, control and Turbine Buildings did not identify any evidence of RCS leakage or equipment damage.

An evaluation of the data from this noise event and the fifth refueling outage noise events (which occurred on January 9, 1992, February 28, 1992 and March 16, 1992) revealed the events had occurred under similar plant conditions. It is believed a small frictional force release in the support system of the RCS occurring during heat-up from cold snutdown is the cause of the event. The exact location of the frictional force release could not be determined. However, analysis of the noise magnitude and consequences demonstrated that the subsequent heat-up and power operation of the Wolf Creek Generating Station (WCGS) would be within the design and licensing bases and would not involve changes in the plant Technical Specifications or the plant license as described in the Updated Safety Analysis Report. Additionally, the magnitude of the noise and the degree of associated RCS movement has reduced in level with each event.

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LICENSEE EVENT REPORT (LER) FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE TEXT CONTINUATION INFORMATION AND RECORDS MANAGEMENT BRANCH (MNSB 7714),- U.S. NUCLEAR REGULATORY COMMISSION, l WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PRO. LECT (3150 0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6: PAGE (3) 05000 TEAR SEQUENTIAL NUMBER REVISION NUMBER HWolf Creek Generating Station 482 93 - - 012 - - 00 2 OF M TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

PLANT CONDITIONS AT TIME OF EVENT

• MODE 3, Hot Standby

• Reactor Coolant System (RCS) at 540 OF

• Check Valve Testing of the Safety Injection Cold and Hot Legs had been completed within the last 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> BASIS FOR REPORTABILITY This Licensee Event Report is being voluntarily submitted to provide the United States Nuclear Regulatory Commission (NRC) with a detailed summary of the Wolf Creek Nuclear Operating Corporation (WCNOC) investigation into the May 10, 1993 noise event.

DESCRIPTION OF EVENT I On May 10, 1993, at 1507 CDT, with the unit in MODE 3, " HOT STANDBY," the Control Room received an alarm on Annunciator 98E, " Seismic Recorder On."

Loose Parts and Vibration Monitoring System (LPVMS) [II] annunciator alarms were also received. Personnel inside the Containment Building

[NH] reported hearing a single loud metallic slap (sound). Personnel in the Auxiliary Building [NF) reported hearing a rumble (thunder like) sound. A Reactor Coolant System (RCS) [AB] leak rate test was performed ,

and an extensive walk down of the RCS, connecting systems, and the Containment, Auxiliary, Control [NA) and Turbine {NM) Buildings did not identify any evidence of RCS leakage or equipment damage resulting from this event.

An evaluation of the data from this noise event and the fifth refueling outage noise events (which occurred on January 9, 1992, February 28, 1992 and March 16, 1992) revealed the events had occurred under similar plant operating conditions. It is believed a small frictional force release in i the support system of the RCS occurring during plant heat-up from cold .i shutdown is the cause of the event. The exact location of the frictional  !

force release could not be determined. However, analysis of the noise magnitude and consequences demonstrated the subsequent heat-up and power operation of the Wolf Creek Generating Station (WCGS) would be within the design and licensing bases and would not involve changes in the plant i Technical Specifications or the plant license as described in the Updated Safety Analysis Report. Additionally, the magnitude of the noise and the associated degree of RCS movement has reduced in level with each subsequent heat up.

i i

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ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THl$

i,iFORMATION COLLECTION REQUEST: 50.0 HRS.

LICENSEE EVENT REPORT (LER) Fo<: WARD COMMENTS RECARDING BURDEN EST! HATE TO THE TEXT CONTINUATION INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMIS$10N, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150 0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER f6: PAGE (3) 05000 YEAR SEQUENTIAL REVISION NUMBER NUMBER Wolf Creek Generating Station 482 93 - ~ 012 - - 00 3N%

TEXT (If more space is required, use additional coptes of NRC Form 366A) (17)

From the fifth refueling outage events, WCNOC Incident Investigation Team (IIT) 92-01 determined that the shims on the RCS crossover leg support structure were binding. This binding allowed a storage of energy which was subsequently released during heat-up resulting in the noise, seismic and LPVMS alarms. The shims were modified during the fifth refueling outage to provide adequate clearance at operating RCS temperature. IIT 92-01 initiated the removal of the shims in sixth refueling outage. IIT 92-01 established commitments to an enhanced RCS thermal growth monitoring program during the next two plant heat-ups from cold shutdown.

The May 10, 1993 event occurred during the first heat-up from cold shutdown after IIT 92-01 was completed. The May 10, 1993 event did not  !

result in the activation of the Operating Basis Earthquake (OBE) Seismic Alarm. Evidence was collected from the seismic instrumentation system, the LPVMS, physical measurements (RCS thermal growth) taken during plant heat-up, field walk downs conducted after the event and the enhanced  ;

thermal monitoring program which consisted of lanyard potentiometers connected to the RCS Crossover Legs to monitor thermal growth in the radial direction. Evidence collected from all of the available methods showed the May 10, 1993 event was similar in nature to but smaller in j magnitude than the 1992 events. Recurrence of the event suggests that j the crossover leg support structure was a major contributor but not the i entire source of the noise that resulted in the sudden release of stored energy from the RCS during the plant heat-up and subsequent expansion.

A safety significance evaluation was performed based upon the evidence  ;

collected and concluded that the noise event did not result in exceeding acceptable safety limits and did not result in any unreviewed safety question. It was determined that subsequent heat-up and power operation of WCGS would be within the design and licensing basis of the plant.

Failure scenarios were developed and root cause evaluation techniques were performed to validate or refute the scenarios. All of the operationally induced failure scenarios were refuted. Thermal expansion growth restraint of the RCS followed by the sudden release of stored energy is the only scenario that could not be refuted and is, therefore, the most probable cause of the event. The location of the restraint has not been conclusively determined, but binding of the reactor vessel nozzle support pads, restraint from insufficient Reactor Coolant Pump

[ABP] (RCP) tie rod support pin clearances, and rubbing at steam generator and RCP articulated vertical supports are considered to be

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FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6: PAGE (3) 05000 YEAR SEQUENTIAL REVISION NUMBER NUMBER Wolf Creek Generating Station 482 93 - 012 - - 00 4 OF 16 1 EXT (if more space is required, use additional copies of NRC Form 366A) (17) l possible contributors. Additional measurements on subsequent plant heat-ups will be required to identify potential locations where significant restraints on thermal expansion motion can occur and verify that adverse conditions leading to the noise event have been eliminated.

DATA COLLECTION ACTIVITIES Physical Measurements RCP tie rod support pin clearances and RCS crossover leg clearances were measured during plant heat-up following the conclusion of the sixth refueling outage. The results of these measurements were compared to the data available from previous plant heat-ups, including the Hot Functional Testing of 1984 and the plant heat-ups of 1992.

The measurements were taken at different temperatures and different l reference points. Therefore, to normalize the data, the overall rate of movement in inches per degree Fahrenheit was determined for each plant plant heat-up for each measured location.

Analysis of measurements of the RCP tie rod support pin clearances during plant heat-up indicated that the support pin clearances on the number 1 and 3 tie rods were closing at a relatively uniform rate as the system temperature increased. The gap on the number 2 tie rods are essentially always closed. This was also observed during the monitored plant heat-up in 1992. The number 2 tie rods on all pumps are oriented approximately perpendicular to a radius from the center of the reactor vessel to the pump. Since the expected thermal expansion motion of the RCS is radially outward from the reactor vessel center line, very little motion or change in the number 2 tie rod gap is expected. The rates of tie rods number 1 and 3 gap closure per degree Fahrenheit change in temperature for the 1993 plant heat-up were essentially the same as for baseline data collected during the hot functional testing in 1984. The tie rod support pin clearances close at a lower temperature than expected by design, but this condition was well documented in 1992 and has not changed.

Measurements of motions of the four crossover legs indicated the following:

1. Movement axially along the crossover leg in the direction of the steam generator was nearly uniform from one plant heat-up to the next.

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FACILITY NAME (1) DOCKET NUMBER (2) M R NUMBER (6: PAGE (3) 05000 YEAR SEQUENTIAL REVISIch NUMBER NUMBER 482 93 - -012 - - 00 5 OF 16 fWolfCreekGeneratingStation TEXT (if more space is required, use additional copies of NRC Form 366A) (17)

I 2. Movement axially along the crossover leg in the direction of the RCPs decreased slightly from the hot functional test on the "A" and "B" RCS loops and remained almost the same on the "C" and "D" RCS loops. This difference is not considered significant.

3. Vertical movement of the crossover leg (downward) for both the steam generator end and the RCP end is nearly identical to that measured >

during hot functional testing.

Field Walk Downs On May 10, 1993, immediately following the event, a walk down was performed on the RCS and the Emergency Core Cooling System (ECCS) The purpose of the walk down was to investigate for signs of damage resulting from the event or conditions that could have contributed to the event.

The scope of the walk down was the RCS piping, including pressurizer, pressurizer spray lines, pressurizer relief lines, surge line and associated supports. A more detailed walk down was performed for the "A" and "D" RCS loops, including attached systems. The walk downs included normal and alternate charging [CB], letdown [CB), main steam (SB], main feed [SJ), accumulator safety injection [ACC], high pressure coolant injection [BQ) , residual heat removal [BO), and steam generator blowdown systems [WI] to the "A" and "D" loops including attached systecs. In addition Reactor Pressure Vessel Torsional Restraints (wagon wheels),

Nuclear Supply Steam System [SB] supports and pipe whip restraints on "A" and "D" loop were included in this walk down.

Similar motion was observed for one of the pins at the top of a vertical support of RCP "A". The motion of this pin also appeared to be quite recent. It is considered that the most probable time for the pin to shift is during a dynamic event, and therefore, it is probable that these pins moved during the event of May 10, 1993. There is evidence that pins on the steam generator and RCP vertical column supports have' shifted in the past. The observed motion of the pins is within the travel allowed  !

I by the pin retaining bolts and the pins are still within the range of positions allowed by the vertical column support design and does not represent a damaged condition of the steam generator vertical support columns. None of the spherical bearings for the column support pinned connections appeared to be worn, shifted, or damaged. The bearings were positioned near the center of the column connecting members as designed.

The movement of the pins is evidence of dynamic motion of the RCS that I

i

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FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6: PAGE (3)

YEAR SEQUENTIAL REVISION 05000 NUMBER NUMBER Wolf Creek Generating Station 482 93 00 6 N 16

- - 012 - -

TEXT (if more space is required, use additional copies of NRC Form 366A) (17) was sufficient to reduce the dead weight loads on the pins and allow them to slide. There was no evidence that the magnitude of the motion was sufficient to damage the vertical column support components.

l The overall evaluation of the walk down was that no degradation of the RC3 or other systems had resulteu from the noise event.

Lanvard Data Collection RCS movement / displacement data was obtained for each RCS loop using lanyard potentiometers to measure horizontal movement perpendicular to the piping connecting the RCP suction to the steam generators. This method of measurement resulted in linear displacement data only. These potentiometers were sampled by a digital data logger at a one second '

rate. The data record consisted of date, time and displacement. The system was calibrated to record directly in inches with a zero indication at the time of installation. The lanyard indicated horizontal movement perpendicular to the crossover piping which approximates the radial direction of growth for the RCS.

The compiled lanyard data showed an increasing deviation of displacement between the locps. Assuming all loops started out with zero displacement ,

in the cold condition, loop "A" and "B" moved the most during the initial plant heat-up with loop "C" lagging far behind. However, during the second plant heat-up period, loop "C" quickly caught up and passed loop "A". On May 11 1993, 1200 CDT, the deviation was 0.152 inches while the total measured d._ 'acement was 1.517 inches (approximately 10% of total travel).

On the surface, this appears as if the RCS is constrained in its growth and residual stresses romain. However, comparing the predicted growth to ,

the average actual growth gives very good correlation within the accuracy of the prediction. The rate of growth at any given time of the various loops was not uniform relative to each other or to the total. However, there were no large cumulative deviations in the displacements. It should also be noted that this data indicates a net displacement of the RCS toward Loops "B" and "C" over the course of the plant heat-up. Since '

each of the lanyards had different reference points on the base of nearby RCP supports, the apparent shift may be somewhat related to thermal expansion of the containment structure due to variations in containment temperature. Comparison of the total diametrical expansions (loops

)

i

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FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER C6: PACE C3)

REVISION 05000 YEAR SEQUENTIAL NUMBER NUMBER OF 16 Wolf Creek Generating Station 482 93 - 012 - - 00 7 ,

TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

"A"+"C" diameter relative to Loops "B"+"D" diameter) showed a differential movement of 0.061 inches just prior to the noise event.

This difference reduced to 0.043 inches immediately after the event and H further decreased to 0.036 inches by 1700 CDT on May 10, 1993 after further adjustment.

Assessment of the data shows movement during plant heat-up was generally smooth and gradual.

In summary, evaluaticn of the lanyard data indicates the overall movement of the RCS is within the expected expansion of the system in the radial direction. No significant anomalies were noted in the overall expansion l of the system although expansion rate and incremental expansion appeared unequal in different directions during various portions of the plant heat-up. Varying restriction of movement at different temperatures between different quadrants of the RCS system resulted in an overall movement in the plant heat-up which continued with resulting changes in direction and movement rates. The RCS appeared to be moving as an entire assembly as the plant heat-up progressed, shifting its centerline relative to the lanyards' reference points. However, overall RCS expansion was continuous. The variations in expansion may be due to an initial "off-centered" position at the beginning of the plant heat-up with the system coming into " centered" position as the RCP tie rods become tight. This would account for the deviati3n in individual loop expansion as the system became " centered",

Personnel Interview and Collection of Process Comouter Records:

During the investigation into the noise event WCNOC interviewed all personnel who were involved in the heatup and / or were in the Containment Building, Auxiliary Building or Control Building to obtain a detailed summary of what they saw, heard and felt. Additionally, pertinent plant process computer records were collected. The performance of the interviews and collection of process computer records was performed to provide support or refute the various scenarios which are discussed in this Licensee Event Report.

Seismic Data:

The magnitude and nature of the event is similar, although less severe, than the occurrences observed in 1992. The responses were well within design limits of the structures and plant safety was never jeopardized.

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FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6: PAGE (3)

YEAR SEQUENTIAL REVISION 05000 NUMBER NUMBER Wolf Creek Generating Station 482 93 - - 012 - - 00 8 OF 16 IEXT (If more space is required, use additional copies of NRC form 366A) (17)

All recorded accelerations were within the Containment Building, similar to the 1992 event.

I The May 10, 1993 event created measurable accelerations similar to those obtained during the noise events that occurred in January, February and March 1992, except the excitations were generally lower in magnitude.

Similarities to the 1992 events:

• All recorded acceleration levels were within the Containment Building and similar in nature to the 1992 event.

i

• The events have the same approximate time duration,

• The east / west response in the reactor cavity is the most prominent.

• The response appears to be initiated in the reactor cavity area.

• Response is similar to that obtained from an instantaneous pulse or shock input.

• The most significant response from the accelerometers was from the one located in the reactor cavity.

• The response was well within design limits: no structural distress is indicated.

Differences from the 1992 events:

Unlike the 1992 events, the excitation inside the containment on May 10, 1993, was not sufficient to trigger the passive seismic instrumentation.

The January 9 and February 28, 1992 events were sufficient to activate the passive instrumentation by exceeding the 32 Hz OBE setpoint level.

However, none of the activation's of the seismic monitoring equipment during the Sixth Refueling Outage activated the passive instrumentation.

-Loose Parts Monitorina System Data:

The behavior of the signals were totally unlike what might be expected from a loose part within the primary system, elimintting a loose part as being a potential cause of the noise.

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FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6: PAGE (3) 05000 TrAR SEQUENTIAL REVISION NUMBER NUMBER g @ jg Wolf Creek Generating Station 482 93 - 012 - 00 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

Similarly the signal behavior is totally different from that exhibited during a check valve test. This eliminates check valve motion as a potential cause of the noise.

• All times are referenced to the reactor vessel. A positive time indicates a later arrival time than at the reactor vessel.

t Temoerature Monitorina Data:

Reactor cavity air temperatures, bioshield concrete temperatures, and containment cooler temperatures were examined for more than a week after <

return to power operations. Evidence the RCS supporting structure continues to plant heat-up for a considerable period of time beyond the cessation of RCS plant heat-up is provided by the concrete temperature indications of the reactor cavity. One week after return to power ,

operation, the bioshield concrete temperature as measured below the RCS piping continued tc increase. Cavity air temperatures were considerably higher than concrete temperatures (approximately 140 OF vs. 108 O F) and h were also increasing but at a slower rate than the concrete, Containment  ;

cooler temperatures (approximately 90 OF) were increasing as well but at a slower rate than the cavity air temperatures. This gives direct evidence the temperatures of the containment structure and RCS supports take a considerable time to reach equilibrium. The dimensional growth of the structures themselves can cause changes sufficient to result in the  :

displacements noted during the 1993 noise event. Additionally, the ambient temperatures are expected tn change with different cooler  ;

lineups, changes in cooling system performance, season, etc. This may  !

result in continuous adjustment of the structure in response to these j temperature changes.

Additionally, the hot leg of the RCS changes temperature in response to reactor power changes. This will result in operationally induced expansion and contraction of the hot Jeg (including resulting movement of the steam Generator and rotation of the crossover leg) during operation, requiring continual adjustment of the RCS with respect to its supports.

Thus, the possibility of " stick-slip" adjustment of the RCS continuing during power operation cannot be discounted. These adjustments may i result in noise and vibration similar to what has already been documented. However, adjustments of this nature are expected to be small (on the order of, or less than the event noted on May 10, 1993) without  !

any significant impact on the RCS or it supports. While each occurrence

9 m

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INFORMATION COLLECTION REQUEST: 50.0 HRS.

LICENSEE EVENT REPORT (LER) FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE TEXT CONTINUATION INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK i REDUCTION FROJErt (3150-0104), OFFICE OF MANAGEMENT AND BUWET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6? PAGE (3) 05000 YEAR SEQUENTIAL REVISION NUMBER NUMBER Wolf Creek Generating Station 482 93 - 012 - - 00 N N TEXT (If more space is required, use additional copies of NRC Form 366A) (17) should be evaluated for its significance and its effect on the life of the plant, there should be no cause for concern and, barring any resulting damage or change in operational characteristics, continued power operation should not be affected.

J FAILURE SCENARIOS EVALUATED, FUTURE MONITORING AND CORR _E_QTIVE ACTIONS FAILURE SCENARIOS:

WCNOC developed seven operationally induced failure scenarios and four

" thermal growth transient failure scenarios" based on information gathered from interviews with plant personnel located throughout the power block, system engineers and the personnel responsible for the check valve monitoring program, plant process computer data, lanyard data, '

manually recorded RCS clearance measurements, field walk downs, related Control Room and surveillance test logs, and standard design documents such as Piping and Instrumentation Diagramc (P&ID's) and Isometric drawings.

The following operationally induced failure scenarios were evaluated and I eliminated as a probable cause for the noise.

1. Main Steam Line water hammer originating outside of containment.

2. Feed Water Check Valve Slam. l

3. Charging / Letdown Transients.  ;

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4. Hydraulic Transient In The Ten Inch Accumulator Line, Cold Leg f I

Injection.

5. Hydraulic Transient In Six Inch RHR Or Two Inch SI Line, Cold Leg Injections.

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6. Hydraulic Transient In One And One Half Inch Boron Injection Tank l (BIT) Line.

7. Hydraulic Transients In 2/6 inch RHR or Si lines, Hot Leg Injection.

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RC M Ib6A U.S. NUCLEAR QEGULATC37 COMMISSION APPROVED B7 OM9 No. 3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. l LICENSEE EVENT REPORT (LER) FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB TEXT CONTINUATION 7714), U.S. NUCLEAR REGULATORY COMMIS$10N, j WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK i REDUCTION PROJECT (3150 0104), OFFICE OF MANAGEMENY AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6: PAGE (3)

REVISION 05000 YEAR SEQUENTIAL NUMBER NUMBER 1 482 M Wolf Creek Generating Station 93 - 012 - - 00 11 TEXT (if more space is required, use additional copies of NRC Form 366A) (17) l THERMAL GROWTH TRANSIENT FAILURE SCENARIOS:

1. Pressurizer Torsional Restraints The pressurizer was considered mainly because it is the only major difference between the loops and because the surge line is connected to Loop D which exhibited the greatest displacement and which is suspected to be the origin of the noise. During inspection of the pressurizer restraints minor discrepancies were also noted. This  ;

scenario considers the binding of the pressurizer torsional restraint as being a possible contributor to the noise event. This scenario involves the sudden release of strain built up between the pressurizer and the Loop D hot leg causing the movements seen on the D Steam Generator vertical support pin and the 15 mil movement of the i crossover leg. This event scenario is considered unlikely because the flexibility of the surge line precludes transmission of sufficient energy to the Reactor Coolant Loops to cause the movement noted in the RCS.

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2. Steam Generator Vertical Surrort Pin ,

The steam generator vertical support columns are connected through a spherical bearing assembly. Along with clearance between the clevis and the ends of the columns, this allows 2 transitional degrees of freedom at the base of the steam generator. As previously discussed, contact was noted between the column and the clevis, Binding at these contact points is considered unlikely because the area over which the restraint could occur was judged to be limited. Interferences noted resulted in a very limited area of contact, and the column has sufficient clearance to rotate away from the interference.

Additionally, at the small areas of contact, the large forces required would result in local deformation rather than significant restraint.

Assuming the bearing is performing its function, the friction in the spherical bearing is considered insufficient to allow significant energy buildup. However, failure of this bearing could be relevant.

No information was obtained during the field walkdowns which would indicate failure of this bearing. Thus, this scenario is considered unlikely to cause the events noted.  ;

3. Steam Generator Lateral Restraint The binding of the Steam Generator lateral restraint was considered one of the possible locations of restrained thermal growth. The steam ,

generator is allowed to grow vertically from the base and to move freely radially from the direction of the reactor vessel centerline.

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kR FGsM 366A U.S. NUCLEAR REGULATOR 7 COMMISSICJ APPROVED B7 OMB NO. 3150-0104 (5 02) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

LICENSEE EVENT REPORT (LER) FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB TEXT CONTINUATION 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6: PAGE (3)

YEAR SEQUENTIAL REVISION 05000 NUMBER NUMBER 12 NM Wolf Creek Generating Station 482 93 - 012 - - oo TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

The steam generator is restrained from movement perpendicular to the radius from the reactor centerline by lateral restraints and the crossover leg. Motion in the radial direction from the reactor centerline is restrained by the Hot leg and by snubbers. Binding at ,

the lateral rest::aints on the Steam Generators showed no evidence of raised metal or any damage to the restraints. Thus, this scenario is considered unlikely to cause the noise events.

4. One or More RCS Nozzle Succort Pads The reactor vessel is supported from attachments on the loop nozzles, two on hot legs and two on cold legs. These nozzles rest on hardened steel plates which are designed to allow radial expansion but to restrict motion. Side plates also restrict translational and rotational motion of the vessel. Since these supports are equally spaced around the vessel, the vessel theoretically should stay in a perfectly ' centered' position at all times. However, due to construction tolerances and variations in heatup of the supports and structures, this cannot be assumed to always hold true. Additionally, friction developed at the sliding surfaces at these supports can be j expected to be a variable and the friction at the supports may not be I identical to each other. This would allow for unequal thermal growth 4 of each nozzle, resulting in buildup of thermal stress. Additionally, unequal friction will allow unequal system expansion.

The support pad of the hot leg connected to Steam Generator D is hypothesized to have the highest friction. The Loop C cold leg pad is  ;

hypothesized to have the lowest friction. During heatup, the reactor, vessel thermally expanded with the majority of motion in the direction of Loop C. The high friction D pad restricted free movement in that l direction. At some high temperature, the gaps in the tie rods on Loop l C closed first, restraining further movement in that direction. With l Loop C motion restrained by the RCP tie rods, thermal stress at the frictional restraint increased faster since further thermal growth was prevented. When the friction was overcome, with the gaps in the RCP tie rods for the other loops almost but not quite closed, the release of energy caused movement in the remaining loops, resulting in an impact when the RCP tie rod gaps closed suddenly. Loop C did nct move because it was already in its proper position. This restraint may come from other locations with the vessel nozzle only one of the possibilities. This scenario is considered the most likely at this time. Future monitoring is expected to pinpoint the location.

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NRt f(CM 366A U.S. NUCLEAR REGULATOQ7 COMM15 Sic 3 APPROVED B7 OMB NO. 3150-0104 (5-02) EXPIRES 5/31/95 l

ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THl$ j INFORMATION COLLECTION REQUEST: 50.0 HRS.  :

LICENSEE EVENT REPORT (LER) FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE j INFORMAfl0N AND RECORDS MANAGEMENT BRANCH (MkBB  ;

TEXT CONTINUATION 7714), U.S. NUCLEAR REGULATORY CCHMIS$10N, j WASHINGTON, DC 20555 0001, AND TO THE PAPERWOf fl REDUCTION PROJECT (3150 0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6: PAGE (3) 05000 YEAR l SEQUENTIAL REvlSION NUMBER NUMBER OF

" Wolf Creek Generating Station 482 93 - 012 - - 00 13 16 TEXT (if more space is required, use additional copies of N*C Form 366A') (17)

H RESULTS:

ROOT CAUSE ANALYSIS:

Formal root cause analysis methods were applied to the failure scenarios l developed. With the available evidence collected, the root cause of the noise event was not able to be determined conclusively. However, the most probable failure scenario of the noise event is restraints on the thermal growth of the RCS followed by sudden release of the elastic energy stored by the restrained motion. The location of the restraint has not been conclusively determined, but binding of the reactor vessel nozzle support pads, restraint from insufficient RCP tie rod support pin clearances, and rubbing at Steam Generator and Reactor Coolant Pumps articulated vertical supports are considered to be likely contributors.

k Evaluation of all evidence concludes the magnitude and nature of the event is similar, although less severe, than the occurrences observed in 1992. The responses were well within design limits of the structures and plant safety was never jeporadized.

SAFETY SIGNIFICANCE EVALUATICN:

The May 10, 1993 noise event did not result in exceeding previously acceptable safety limits and did not result in any unreviewed safety questions. Subsequent plant heat-up and power operation of the WCGS will be within the design and licensing bases. This conclusion is based on the following:

1. It has been demonstrated that the loads and stresses in the piping, supports, and primary equipment nozzles remain within the appropriate allowable values. The functions of the RCS, engineered safety features actuation system, emergency core cooling system, and attached auxiliary systems are maintained. Therefore the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased. The occurrence of the noise event during the sixth refueling outage does not adversely affect the fuel integrity since it does not create a condition where any fuel design criteria are exceeded. It is concluded that the May 10, 1993 noise event had no adverse effect on the safety analysis for the WCGS or on the potential for failure of equipment important to safety. Therefore, no

O NEC FORM 366A U.S. NUCLEAR DEGULATCW COMMISSIC3 APPR0"ED BY OMB NO. 3150-0104

2) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS ,

INFORMATION COLLECTION REQUEST: 50.0 HRS.

LICENSEE EVENT REPORT (LER) FORWARD COMMENTS REGARDING BURDEN ESTIMATE 10 THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB TEXT CONTINUATION 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHI NGT ON, DC 20555 0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6: PAGE (3)

YEAR SEQUENTIAL REVISION 05000 NUMBER NUMBER OF 16

- 012 - - 00 Wolf Creek Generating Station 482 93 14 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) additional accident scenarios which may result in the failure of  ;

equipment are created.

2. Continued plant heat-up and subsequent power operation of WCGS following the May 10, 1993 noise event does not create the possibility for an accident or equipment malfunction of a different type than any previously evaluated in the USAR. No new failure modes or limiting ,

single failures were created as a result of the May 10, 1993 noise event. Therefore, no new accident scenarios or equipment malfunctions need be considered.

3. Continued plant heat-up and subsequent power operation of WCGS following the May 10, 1993 noise event does not reduce the margin of safety as defined in the Bases of the WCGS Technical Specifications.

The margin of safety of the RCS piping, components, and supports is f

defined by the structural criteria presented in Section III of the ASME Boiler and Pressure Vessel Code. Considering the current configurations of the affected systems, as well as the potential loading experienced during the noise phenomenon, it has been y demonstrated that the appropriate criteria were met and that the I functions of these systems are maintained.

The WCGS Technical Specifications ensure that the plant operates in a l manner that provides an acceptable level of protection for the health and l 1

safety of the public. The WCGS Technical Specifications are based upon assumptions made in the safety and accident analyses, including those relating to the RCS structural design. This ensures adequate margin to the regulated acceptance criteria for the accident analyses. Since it has been concluded the the RCS and associated auxiliary system design l parameters and assumptions which go into the accident analyses remain appropriate, the conclusions in the USAR remain valid. Therefore, the ,

regulated margin of safety as defined in the Technical Specifications is not affected. l 1

1 The investigation established that the noises were not the result of a loose part within the RCS by satisfactorily completing system integrity  !

inspections and analyzing LPVMS data. WCGS also verified that the core l remained intact by exercising all control rods as part of the normal l plant startup testing. Additionally the core flux mapping system in-core probes were exercised during power ascension. The structural integrity of the fuel is designed for lateral non-operational loads and normal and abnormal vent loads. The non-operational design loads are based on 4 g l l

kRC FORM 366A U.S. NUCLEAR REGULATC37 COMMISSICD APPROVED BY OMB No. 3150-0104 (542) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

LICENSEE EVENT REPORT (LER) FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB TEXT CONTINUATION 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTI(N PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY N 9E (1) DOCKET NUMBER (2) LER NUMBER (65 PAGE (3) _

05000 TEAR SEQUENTIAL REVISION N'JMBER NUMBER 6NN Wolf Creek Generating Station 482 93 012 -- 00 TEXT (if mote space is required, use additional copies of ORC Form 366A) (17) axial and 6 g 1.ateral accelerations. The abnormal design loads are based on a combination of Safe Shutdcen Tarthquake (SSE) and blowdown forces from a worst case Loss of Coolant Accident. The licensing basis SSE maximum ground acceleration is 0.20 g. The events recorded during the fifth and sixth refuelir.g outages exhibited accelerations for the reactor cavity area significantly less than design values. Additionally, visual inspections conducted an the lower sections of the fuel assemblies during the sixth refueling outage did not identify any structural damage to the fuel as a result of any of the fifth refueling outage noise events.

Therefore core integrity has been maintained.

CORRECTIVE ACTIONS:

An Enhanced Thermal Expansion Monitoring Program will be developed and implemented during future plant heatup and cooldown sequences until further benefit is deemed negligible. The purposes of this monitoring program will be to:

• Aid in the identification of locations of restraint on RCS thermal growth.

• Develop information regarding the energy storage and release mechanisms that will confirm restrained growth or release events cannot reach damaging magnitudes.

• Verify that implemented corrective actions have been as effective as expected.

The Thermal Expansion Monitoring Equipment will be insLilled prior to Refueling VII.

PREVIOUS SIMILAR EVENTS:

Licensee Event Report 92-006-00 identified ano discussed the the January, February and March 1992 noise events at the Wolf Creek Generating Station. The events occurred under similar plant conditions as the M v 10, 1993 noise event. Based on results from the Incident Investigation Team formed to evaluate these events it was determined that the shims on i the RCS crossover leg support structure were binding. This binding allowed a storage of energy which was subsequently resleased during plant heat-up resulting in the noise, seismic and LPVMS alarms. The shims were

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NRC.rcaM 366A U.S. NUCLEAR REGULATC3Y COMMISSICC APPROVED B7 OMB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

LICENSEE EVENT REPORT (LER) FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB TEXT CONTINUATION 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF ,

MANAGEMENT AND BUDGET, WASHlWGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6: PAGE (3) 05000 YEAR SEQUENTIAL REVISION NUMBER NUMBER Wolf Creek Generating Station 482 93 - 012 - - 00  %@ M TERT (If more space is required, use additional copies of NRC Form 366A) (17) modified to provide adequate clearance at operating RCS temperature The shims were removed during the sixth refueling outage. Evidence collected from t.1 of the available methods showed the May 10, 1993 event was similar in nature to but smaller in samller in magnitude than the 1992 events. Recurrence of the event suggests that the crossover leg support stri.cture was a major contributor to but not the entire source of the noise that resulted in the sudden release of stored energy from the RCS during the plant heat-up and subsequent expansion.

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