ML20137G577
| ML20137G577 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 03/24/1997 |
| From: | Dacimo F, Pfister E COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LER-970324, NUDOCS 9704010433 | |
| Download: ML20137G577 (6) | |
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C.ommonwealth litiwn Company '
IMalle Generating $tation
.O 2601 North 21st Road Mar *.ettles,.11. 613 i 1 9757 '
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Tel 815 357&61 l
March 24,1997 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 Licensee Event Report #97-008-00, Docket #050-373 is being submitted to your l
office in accordance with 10 CFR 50.73(a)(2)(i).
t Respectfuily, j
i r
f Fr(d Dacimo Plan, General Manager 4
LaSalle County Station Enciosure cc:
A. B. Beach, NRC Region 111 Administrator M. P. Huur, NRC Senior Resident inspector - LaSalle C. H. Mathews, IDNS Resident inspector - LaSalle I
F. Niziolek, IDNS Senior Reactor Analyst INPO - Records Center i
9704010433 970324 j(h-f l PDR ADOCK 05000373 I
S PDR 01000F 15p55555P58 me.m o.mn.,,
NRC FORh1366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (5-92)
EXPIRES 05/31/95 ESTIM ATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER)
COMMENTS REGARDING BURDEN ESTIMA~t TO THE l
INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 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) 05000373 PAGE (3)
LaSalle County Station Unit One 1 of 5 1 TITLE (4)
Rod Block Monitor Not Enabled At Power Level Required By Technical Specifications Due To Nonconservative Test Method In An Inadequate Procedure l
EVENT DATE 151 l _LER NUMBER (6) l REPORT DATE (7) l OTIIER FACILITIFS INVOLVED (8)
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FACILITY NAME DOCKLT NUMBER mun scMnLa LaSalle County Station 05000374 Unit Two 02 20 97 97 008 00 03 24 97 FACILITY NAME DOCKET NUMBER
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NRC Fonn 366A) 7 m..<a t ICENSEE CONTACT FOR TIIIS LER U 2)
NAME TELEPHONE NUMJTR (include Area Code)
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l Eugene Pfister, System Engineer (815) 357-6761 Extension 2740 i
COMPLETE ONE I INE FOR EACH COMPONENT FAILURE DESCRIHED IN THIS REPORT (13)
(.AUSL Sv51LM COMPUNLN T MANUF AcrLRLk (AUS
$) S'lLM COMPONEN1 M ANUF AC1LkLk RLK*I ABLE TO NPRDS TO NPRDS j
SUPPLEMENTAL REPORT EXPFCTED 04 EXPECTED MONT H DAY vlAk YES NO SUBMISSION (if yes, complete EXPECD SUBMISSION DATE)
DATE (15)
AHSTR ACT (Lumt to 1400 spaces,i c., approximately fifteen single-space typewntten hnes 16)
On February 20, 1997, a determination was made that calibration procedures for the Unit 1 and 2 Rod Block Monitor (RBM) did not verify that the RBM would be enabled at prior to obtaining 30 percent Core Thermal Power (CTP) as required by plant Technical Specifications.
This limit is established to ensure adequate protection against control rod withdrawal errors at high power.
Subsequent investigation determined, on February 26, 1997, that on 4 occasions between 1988 and 1993 the RBM reference downscale bypass functions were calibrated in the decreasing power direction with setpoints which may have resulted in the RBM being enabled between 30.2 and 31.68 percent CTP.
This was due to a hysteresis (reset differential) of as much as 2.7 percent between the decreasing and increasing setpoints.
Due to Critical Power Ratio (CPR) margin present during low speed recirculation pump operation, there would have been no challenge of the Minimum Critical Power Ratio (MCPR) safety limit had a control rod withdrawal error event taken place.
Thus, the safety significance of this event was minimal.
This event was caused by inadequate procedures. The affected procedures and instrument data sheets will be revised.
Selected calibration procedures will be reviewed for similar problems.
)
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NRC FOKht 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (5 92)
EXPIRES 05/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS I
j INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER)
COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH TEXT CONTINUATION (MNBB 7734). U.S. NUCLEAR REGULATORY COMMIS$10N, j
WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF M AN AGEMENT AND j
BUDGET, WASHINGTON. DC 20503.
FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
YLAR bbQUEN11AL REVISION NUMBER NUMBER j
LaSalle County Station Unit one 05000373 97 008 00 2 of 5 (If more space is required, use additional copies of NRC Forrn 366A)(17)
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i PLANT AND SYSTEM IDENTIFICATION
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General Electric - Boiling Water Reactor i
Energy Industry Identification System (EIIS) codes are identified in the text as f
(XX).
l 4
A.
CONDITION PRIOR TO EVENT I
Unit (s): 1/2 Event Date: 02/20/97 Event Time:
1400 Hours i
Reactor Mode (s): 4/N Mode (s) Name: Cold Power Level (s): 0%/0%
Shutdown /Defueled i
S.
DESCRIPTION OF EVENT j
On February 20, 1997, a determination was made that calibration procedures for the Unit 1 and Unit 2 Rod Block Monitor (RBM) [IG) did not verify that the RBM I
j would be enabled prior to obtaining 30 percent CTP as required by plant Technical Specifications. This discovery occurred during a system functional review of the Neutron Monitoring System when these instrument calibration procedures where checked against Operational Experience (OE) item 8225.
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Plant Technical Specifications and UFSAR require the RBM to be operable at thermal power levels of >/= 30 percent of rated.
The RBM is automatically
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bypassed when a reference power level signal from the Average Power Range Monitor (APRM) Systen. decreases below the reference downscale setpoint.
l This setpoint is specified in calibration procedures LIS-NR-105 and LIS-NR-205 as j
27 +/- 0.25 percent CTP.
However, the setpoint is only measured in the decreasing power direction.
It is the increasing setpoint which is important to j
enabling the RBM prior to the point where it is required to provide protection i
against rod withdrawal errors at high power levels. Neither the increasing setpoint nor the electronic hysteresis (reset differential) of the trip circuit is measured by the calibration procedures.
Thus, the nonconservative test method in these procedures provides no assurance that the RBM is enabled prior to exceeding 30 percent CTP.
At the time of discovery, Problem Identification Form (PIF) 97-1253 was initiated to review past calibration results to determine if the increasing setpoint had ever been greater than 30 percent CTP.
Both RBM channels on Units 1 and 2 were declared inoperable pending performance of a revised calibration procedure. No additional actions were required since the RBM channels were not required to be operable with Unit 1 in Operational Condition 4, Cold Shutdown, and Unit 2 defueled.
To establish a value for the electronic hysteresis, tests were performed on 16 trip comparators of the type used for the RBM reference downscale function.
The tests found an average hysteresis of 1.93 percent and established a 99.5 percent statistical confidence level that the hysteresis would not exceed 2.7 percent.
A review of the calibration procedure revision history determined that the current method for measuring this setpoint has existed, without change, since the original preparation of the procedure. However, prior to July, 1988, a setpoint of 20 percent CTP was used for the reference downscale function.
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NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104
{
(5-92)
EXPIRES 05/31/95 1
ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLIEUlON REQUEST: 50.0 HRS. FORWARD i
LICENSEE EVENT REPORT (LER)
COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH TEXT CONTINUATION (MNBB 7714). U.S. NUCLEAR REGULATORY COMMISSION, j
j WASil!NGTON. DC 20555-0001. AND TO THE PAPERWORK 1
REDUCTION PROJFC (3150-0104) OFFICE OF M ANAGEMENT AND j
BUDGET, WASHINGTON, DC 20501 i
FACII ITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
YtAR bl@ULN IlAL REV1 SKIN j
NUMBfR Nt 'M R FR 3
LaSalle County Station Unit One 05000373 97 008 00 3 of 5 (if more space is required, use additional copies of NRC Form 366A)(17) 4 l
The as found setpoint values for calibrations performed between July, 1988 and i
present were reviewed. To these values to provide an estimate of the increasing
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power setpoints, 2.7 percent was added. Due to the large amount of conservatism j
in the setpoint prior to July, 1988, a review of these calibration results was not deemed necessary.
4 1
On February 26, 1997 at 1130, the setpoint review established a significant potential for nonconservative RBM reference downscale setpoints on 4 occasions.
j The dates of these occurrences were December 13, 1988 (Unit 1, RBM B),
1 August 21, 1990 (Unit 2, RBM B), June 10, 1992 (Unit 2, RBM B) and August 4, 1993 j
(Unit 1 RBM A & B). The calculated as found increasing setpoints for these i
occurrences ranged from 30.2 percent to 31.69 percent CTP.
The values in excess i
of 30 percent CTP were due to setpoint drift, within normal limits, and the previously unaccounted for hysteresis.
PIF 97-1491 was written to document this finding and initiate the LER process.
During all except the August 4, 1993 occurrence, only one of the two, redundant RBM channels was affected.
Based on a review of plant work history, the other channel remained operable during these time periods and would have performed its intended function.
During the periods of concern for Unit 1, the unit operated continuously at power levels greater thar. 35 percent CTP.
Thus, operation was at power levels above the calculated increasing setpoints and there is reasonable assurance that both RBMs were operable.
For Unit 2, several startups and shutdowns occurred during both of the titae
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periods of concern.
However, during each startup, the Unit operator is required
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by procedure to verify that both RBM channels are enabled at 30 percent CTP.
Either APRM indication or heat balance is used for this determination.
Since this verification was performed and the calculated increasing setpoints were only 30.33 percent CTP for both occurrences, it is probable that the actual hysteresis was less than the assumed 2.7 percent value and the RBM was enabled at
>/= 30 percent CTP.
This condition is reportable per 10 CFR 50.73 (a) (2) (1) (B) due to a condition prohibited by the plant's Technical Specifications.
C.
CAUSE OF EVENT This event was caused by an inadequate calibration procedure.
The reason for performing the setpoint measurement in the decreasing direction could not be determined from the documentation of the original calibration procedure draft in i
1977.
Likewise, the documentation for the 1988 procedure revision which changed the setpoint from 20 percent to 27 percent CTP did not mention any considerations j
related to this issue other than the actual setpoint change.
Other procedure revisions addressed technical and performance issues not related to this setpoint or function.
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NRC FORpt 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (5 92)
EXPIRES 05/31/95 4
4 ESTIMATED BURDEN PER RESPONSE TO COMPL.Y WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FO4 WARD LICENSEE EVENT REPORT (LER)
COMMENTS REGARDING BURDEN ESTIMATE TO THE J
INFORMATION AND RECORDS MANAGEMENT BRANCH TIiXT CONTINUATION (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION,
}
WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK 1~
l REDUCTION PROJECT (3150-0104), OFFICE OF M ANAGEMENT AND BUDGET, WASHINGTON, DC 20501 FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
)
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RAk SiglLNTIAL ktVlilON 1
Nt HRF R Nt'MBF R j
LaSalle County Station Unit One 05000373 97 000 00 4 of 5
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(if more space is required, use additional copics of NRC Form 366A)(17)
D.
ASSESSMENT OF SAFETY CONSEQUENCES j
The safety function of the RBM is to mitigate the consequences of a Rod i
Withdrawal Error (RWE) by initiating a control rod withdrawal block. During a RWE, an out of sequence control rod is selected and continuously withdrawn. The j
localized increase in power causes the surrounding fuel pins to approach the j
point of departure from nucleate boiling (DNB). The RBM uses tha Local Power Range Monitors (LPRMs) to detect the localized power increase and terminate the l
rod withdrawal prior to violating the Minimum Critical Power Ratio (MCPR) safety limit. The MCPR safety limit ensures less than 0.1 percent of the fuel pins l
experience DNB.
3 i
During the cycle transient analysis, the setpoint for which the RBM should block
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control rod withdrawal is determined based on the full power MCPR margin desired.
As the setpoint increases, so does the margin needed at the initial condition to i
the transient. Recent transient analysis assumes the rod withdrawal is not j
stopped during the transient, since current core designs do not require the 1
additional MCPR margin.
Thus, the cycle specific CPR margin for the RWE I
transient assumes no credit for the RBM mitigating the consequences.
For the i
four occurrences when the bypass setpoint was found out of tolerance, however, i
the RWE delta CPR was based on the RBM terminating the event when the localized j
power approached a setpoint of 110 percent.
During the four periods above, this j
function may not have occurred between 30 and 35 percent CTP.
During these times, however, more than adequate delta CPR margin was present since reactor I
recirculation pumps were in slow speed with testal core flow less than 40 percent.
l For those instances, addition CPR margin was gained in the form of the Kf factor.
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1 This is a flow dependent multiplier to the CPR limit that maintains increased margin at reduced flows for the reactor recirculation flow runup transient. This i
margin more than adequately covers the difference in delta CPR between a full power RWE with a 110 percent rod block, and an RWE unblocked.
l i
It is therefore concluded that the error in calibrating the bypass setpoint had minimal effects on plant safety.
Since the setpoint may have only been bypassed 1
between 30 and 35 percent CTP, and additional delta CPR margin is present in this j
condition, the MCPR safety limit was never challenged, i
E.
CORRECTIVE ACTIONS I
t 1.
Calibration procedures LIS-NR-105 and LIS-NR-205 will be revised, prior to c
operation of the associated unit above 30 percent CTP, to incorporate the correct RBM reference downscale test method. Testing the setpoint in the increasing direction will ensure that the RBM is enabled prior to 30 percent CTP.
The hysteresis will then be in the decreasing direction causing the RBM bypass to occur at a lower CTP.
2.
RBM calibrations using the revised procedures will be performed in accordance with the existing surveillance schedule prior to operation of the associated unit above 30 percent CTP.
3.
Prior to startup of the associated unit, a note will be added to the RBM instrument data sheets to emphasize that the reference downscale must be tested in the increasing direction.
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NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (5-92)
EXPIRES 05/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS t
INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER)
COMMENTS REGARDING BURDEN ESTIM ATE TO THE 4
INFORMATlON AND RECORDS MANAGE:alENT BRANCH I
TEXT CONTINUATION (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASillNGTON, DC 20555-0001, ANDTOTilEPAPERWORK REDUCTION PROJECT (3150-0104) OWICE OF MANAGEMENT AND BUDGET, WASHINGTON. DC 20503.
FACILITY NAME (1)
DOCKFT NUMBER (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL REYlSION NUMRf'R NUMHf'R LaSalle County Station Unit One 05000373 97 008 00 5 of 5
)
(If more space is required, use additional copies of NRC Form 366A)(17) 4.
A System Functional Review Program, currently in progress, provides added confidence that the Technical Specifications are consistent with other plant i
documentation and/or are being appropriately implemented. This program was I
initiated to establish a level of confidence that selected systems demonstrate i
performance consistent with the desiCn basis. One of the elements of the program is to identify required system functions and sub-functions as described in design bases documents including the Technical Specifications.
Surveillance testing requirements and procedures and other test documentation are then reviewed to confirm that system functionality is demonstrated. As stated previously, it is this program which identified the aforementioned discrepency. Any inconsistencies identified among the source documents are being documented and tracked to resolution. This program is being implemented with applicable corrective actions completed prior to restart.
F.
PREVIOUS OCCURRENCES LER NUMBER TITLE 373-89-013 Suppression Chamber High Level Alarm Exceeded Due to Procedural Inadequacy Corrective actions of this LER addressed only the associated procedure and instrumentation. No review was done to identify similar deficiencies in other procedures.
G.
COMPONENT FAILURE DATA
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Since no component failure occurred, this section is not applicable.