05000341/LER-1995-001, :on 950211,wide Range Reactor Water Level Recorder Indicated 10 Inch Reactor Water Level Decreased Due to Inadequate Test Controls.Controlled Diagnostic Monitoring Instruments to Prevent Similar Conditions
| ML20081C134 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 03/13/1995 |
| From: | Martin J DETROIT EDISON CO. |
| To: | |
| Shared Package | |
| ML20081C111 | List:
|
| References | |
| LER-95-001, LER-95-1, NUDOCS 9503170130 | |
| Download: ML20081C134 (9) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(x) |
| 3411995001R00 - NRC Website | |
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FOCILITY NAME (1)
DOCKET NUMBER (2)
PAGE (3)
Fermi 2 05000 341 1 OF 8 TITLE (4)
Reactor Water Level Indication Transient EVENT DATE (5)
LER NUMBER (6h REPORT NUMBER (7)
OTHER FACILITIES INVOLVED (8)
F ACsuTV NAME DOCKE'i NUM3ER SEQUENT!AL REVISION MNH AY YEAR YEAR MN n
YEAR NUMBER NUMBER 05000 F ACluTY NAME DOCKET NUMBER 02 11 95 95 001
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00 03 13 95 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 4: (Check one or more) (11)
MODE (9) 1 20 402(b) 20.405(c) 50.73(a)(2)(iv) 73.71(b)
POWER 20 405ta)(1)(1) 50.36(c)(1) 50.73(a)(2)(v) 73.71lc)
LEVEL (10) 009 20.405(a)(1)(ii)
Sc aste)(2) y 50.73(a)(2)(vii)
OTHER 20.405(a)(1){ni) 50.73(a)(2)(i) 50.73(a)(2)(viii)(A)
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Form 366A) 20 405(s)(1)(v) 50.73(a)(2)(iii) 50.73(a)(2)(x)
LICENSEE CONTACT FOR THIS LER (12)
NAME TELEPHONE NUMBLH (irmiuoe Area Coce)
Jimny L. Martin, CoTpliance Engin r (313)586-4225 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
CAUSE
SYSTEM COMPONE NT MANur ACTURE R CAUS SYS M COMPONENT MANUF ACTURER O PR S SUPPLEMENTAL REPORT EXPECTED (14)
EXPECTED MONTH DAY YEAR YEs SUBMISSION p yes. covee ExprCTED sueMiss>ON DArEi X
OATE (15)
ABSTRACT (Limit to 1400 spaces, i e., approximately 15 single-spaced typewritten lines) (16)
On February 11,1995 at 1718 hours0.0199 days <br />0.477 hours <br />0.00284 weeks <br />6.53699e-4 months <br />, Control Room operators noticed that the wide range reactor water level recorder indicated a 10 inch reactor water level decrease. This crTor in level indication occurred when diagnostic monitoring instmmentation was manually de-energized while connected to the affected plant process instrumentation on the Division I and Division 2 testability panels. The diagnostic monitoring instruments were disconnected and the wide range level indications were restored to normal within approximately 40 minutes. The narrow range level and pressure instruments were momentarily affected during the process of disconnecting the diagnostic monitoring instruments. A review of the affected Reactor Pressure Vessel (RPV)
Water 1.evel and Pressure Instrumentation determined that the level 8 trips for High Pressure Coolant Injection and Reactor Core Isolation Cooling Systems would have been non-conservative due to this level shift. This event occurred due to inadequate test controls and because the input impedance characteristic of the diagnostic monitoring instrumentation changed unexpectedly when the equipment was de-energized. This impedance characteristic was not recognized by the vendor or site testing personnel prior to use. Corrective actions include controlling the diagnostic monitoring instruments to prevent similar conditions, evaluating industry good practices for test control and the use of monitoring and test equipment to identify any necessary programmatic improvements, revising procedures and providing lessons learned training.
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EVENT DATE 2 PER BLOCK 7 TOTAL
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1 OPERATING MODE 10 3
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Initial Plant Condition:
Operational Condition:
1 (Run)
Reactor Power:
9 Percent Reactor Pressure:
950 psig Reactor Temperature:
540 degrees Fahrenheit l
l Descriotion of the Event:
l l
During an investigation to determine the cause of spurious RCIC (BN) level 8 alarms received on main turbine trips, a work request was issued to acquire reactor water level and pressure instrumentation response information for transient monitoring during turbine testing. The j
monitoring equipment consisted of two Diagnostic Data Acquisition System (DAS) computers, connected to non safety related power supplies, whose input terminals were connected in parallel with the inputs of the Division 1 and Division 2 instrumentation at the testability panels at various times over the period of February 10 and 11,1995. One DAS unit was connected to instruments in panels H21P082 and H21P084 and the other unit was connected to instruments in panels H21P083 and H21P085 as described below.
Panels H21P082 and H21P084 B21N680A - Reactor Water Level, Narrow Range, (RPS) Level L3 B21N695C - Reactor Water Level, Narrow Range, Main Turbine and Feedwater Pump, Level L8 b21N691 A,C - Reactor Water level, Wide Range (ECCS), Levels L1, L2 and L8 B21N690A - Reactor Pressure ECCS, LPCI and Core Spray Injection Panels H21P083 and H21P085 B21N680C - Reactor Water level, Narrow Range, (RPS) Level L3 B21N695D - Reactor Water level, Narrow Range, Main Turbine and Feedwater Pump, Level L8 B21N691B D - Reactor Water Level, Wide Range (ECCS), Levels L1, L2 and L8 B21N690B - Reactor Pressure, ECCS, LPCI and Core Spray Injection l
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,o.c.,, a.,,,.w..~.- n, emum o n During a break in turbine testing activities on February 11,1995, at approximately 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br />, the DAS units were manually de-energized but left connected to the testjacks on the testability panels. Approximately 5 minutes after the DAS units were de-energized, operators (utility, licensed) observed the reactor vessel level indication on the wide range recorder in the Main Control Room indicating a reactor water level ten (10) inches below the level shown on the nanow range indicators. Instmment and Control Shop personnel (utility, non-licensed) were contacted and the DAS units were disconnected from the testability panels. The Wide Range Reactor Water Levelindication returned to the proper indication once the DAS units had been disconnected.
Momentary effects were observed on the connected reactor pressure and narrow range reactor water level instmmentation (PI)(LI) while the DAS units were being disconnected. First the Division I Wide Range Reactor Water Level loops were disconnected from the DAS unit, followed by the Natrow Range Reactor Water Level loops connected to that unit, then the Division II Wide Range loops were disconnected from the second DAS unit, followed by the associated Narrow Range loops; then the Division II pressure loop was disconnected, followed by the Division I pressure loop. When the Wide Range loops were disconnected, the over-i voltage diodes connected to the Narrow Range and Pressure loops became forward biased. This resulted in some of the transmitters
- current being shunted through the DAS. Modeling predicts this could result in about a 2 inch drop in Nanow Range indication and about a 40 psi drop in pressure indication. However, disconnecting the Narrow Range loops would have had no additional effect on the pressure instruments.
Subsequent to the event, the vendor was contacted for additional information on the DAS instmments and an analysis was performed by Plant Engineering (utility, non-licensed). The engineering analysis of the DAS instruments revealed that when the DAS is de-energized, the inputs are connected together through voltage clamping circuits in solid-state multiplexer integrated circuits. Normal input voltages from the Testability Trip Unit forward biased the clamping diodes in these integrated circuits resulting in excessive current being drawn from the circuit under test. The amount of cunent drawn from the circuit under test varies due to the nonlinerarities of the clamping diodes and the DAS power supply. As a result, the error in wide range level indication would increase to about 17 inches if the indicated level increased to 214 i
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inches (setpoint for high level, L8, trips). The Wide Range Reactor level loop current caused a l
voltage drop across the effective resistance of the DAS power supply that kept the over voltage i
diodes connected to the Reactor Pressure and Narrow Range Reactor Level Loops from being I
tumed on. As a result, as long as the wide range reactor level loops were connected to the DAS, there was no effect on the narrow range reactor level instmments.
A review of the affected Reactor Water Level and Pressure Instrumentation Channels predicted f
that the Level 8 trip from HPCI(BJ) and RCIC would be non-conservative. It was predicted that j
a Level 8 trip of HPCI and RCIC would have occurred at a higher level than the Technical Specification allowable value of 219 inches.
i Cause of the Event-l l
This event occurred due to inadequate test controls resulting in an inadequate assessment of the potential effect of the DAS used to monitor safety-related instrumentation. Prior to use, the input impedance was measured on the DAS with a digital voltmeter and found to be acceptable.
However, the evaluation of the DAS equipment prior to use did not apply a test voltage to simulate conditions that would be seen in the field (1 to 5 volts). The vendor manualincluded information regarding input impedance but provided no caution of the low input impedance when the DAS was switched off. This contributed to an incorrect assumption that the equipment was non-intrusive.
The connection of the DAS equipment should have been controlled under the temporary modification procedure. Guidance in the temporary modification procedure did not distinguish between connection of diagnostic equipment, which should be govemed by the procedure, and i
taking momentary readings with test instruments. As a result, the distinction between these types ofjobs was unclear. This had led to a general practice of considering high impedance diagnostic equipment as non intrusive; thus temporary modifications were not being used for these installations. Had the temporary modification process been used, the need for a safety evaluation would have been considered, and may have prevented simultaneously connecting multiple channels within a division to the same computer and connecting both divisions to DAS units at j
the same time.
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Analysis of the Event
As previously discussed, Reactor Vessel Water Level Instrumentation is provided to monitor the entire range of operation over four zones of measurement. These ranges are Narrow Range, Wide Range, Core Level and Flood-up Level. Only some of the Narrow Range and Wide Range Level instmments were affected during the testing activity.
The Narrow and Wide Range instruments provide level measurement for safety functions including RPS OC) and ECCS initiations at various water level setpoints, i.e. L8, L3, L2 and Ll.
Manual shutdown of the DAS units while the test leads remained connected resulted in a decrease in indicated wide range level of approximately 10 inches at the normal level. The same effects would have occurred with the loss of station or offsite power.
The inaccuracy imposed by the diagnostic monitoring equipment would not have prevented the low level trips. The negative level bias introduced by the DAS is almost negligible at low water level and is in the conservative direction with regards to low level trips and actuations. However, because the effect of bias resulted in lower level as indicated on the wide range instrumentation, the HPCI and RCIC systems Level 8 trip would have occurred at a higher actual level than expected.
Level 8 trip signals indicate that the reactor water level has increased and that protective actions are required to prevent overfilling the reactor vessel. The trip signals are selected low enough to protect all of the steam turbines against gross carryover of moisture and to provide adequate core thermal margins during abnormal events.
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NJMBER NUMBER Fermi 2 05000 341 6
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un m n..,n.a,..au.,m o m Cr.,um on Functions activated with Level 8 signals are generated from two different reactor water level measurement systems. They are the narrow range, which has a span of 60 inches and the wide range, which has a span of 210 inches. The narrow range generates the trip signals for the main turbine and feedwater pump turbines. The wide range generates shutdown signals for the RCIC and HPCI systems.
The allowable value (AV) < 219 inches for the Level 8 trip setpoint was established based on the most severe applicable event, i.e. the feedwater controller failure that can directly cause an increase in the feedwater flow. (Note: All water levels are given with respect to the distance above the top of the active fuel). The AV is specified at approximately the top of the steam separators near the upper limit of the region where acceptable steam carryovers occur. Because the DAS changed the indicated level of wide range channels only, the safety design basis (i.e.,
the minimum critical power ratio (MCPR)], would not have been affected during the postulated transient.
If a small break Loss of Coolant Accident is assumed, the reactor scrams as expected at low level (L3), or on high drywell pressure and L2 actions occur earlier than expected, e.g. the HPCI and RCIC initiation. With an estimated 17 inch bias on the wide range instrument at the level 8 trip setpoint, the level inside and outside the dryer seal skirt will reach about 231 inches (L8 Nominal l
Trip Setpoint of 214 inches + 17 inches) before the delayed HPCI and RCIC trip would occur.
This trip setpoint is approaching, but still below the upper instrument nozzle; therefore the trips would be expected to function.
However, even at this higher water level, the peak level would remain well below the elevntion of the bottom of the main steain lines. No significant amount of moisture carryover would occur because the power decrease due to a scram early in the event would unload the steam separators and dryers so that they could easily handle the conditions near the high level trips. The trip of the feedwater pump turbines would occur as expected at the actual Level 8 setpoint since this function is actuated from the narrow range instruments which were working properly.
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Since the steam lines would have remained clear, the RCIC and/or HPCI systems would have l
remained available if they were needed for subsequent inventory supply. Additionally,if the control room operators were aware of the water level anomalies, their actions in such a situation would be to take control of systems feeding the vessel to restore and maintain normal water level.
If the negative level bias was assumed to affect the pressure and narrow range level instruments (when the DAS was disconnected from the wide range), the limiting transient.would be the failure of the feedwater controller to the maximuu demand. In this transient the feedwater pump turbine (only one feedwater pump was in operation at a such low power level) trip would have occurred at a higher actual level than expected (about 17 inches above the L8 Nominal Trip Setpoint of 214 inches). The main turbine was off line at the low power level that existed at the time of this event,i.e. less than 10% power. Even though the feedwater pump turbine trips would have occurred above the AV, the trips would have stopped the water level increase well below the elevation of the main steam lines so that no water would have carried over. Due to the very large initial thermal margin at such a low initial power level, the fuel thermal margins would have remained well within the safety limits.
It is concluded that a 10 inch bias (or 17 inches at the actual L8) of the Fermi 2 water level instruments would not cause excessive moisture or flooding of the main steam line (even for the conservative case postulated), and therefore the RCIC and/or HPCI Systems would remain available to restart if neede i for longer term inventory supply.
The affected pressure instruments provide input to the low pressure injection permissive for LPCI (BO) and Core Spray (BM) injection valves. The downward bias would have allowed the affected channels to trip early (at slightly high pressure). However, this would have had no effect on the operation of the injection valves because the logic also requires low pressure trips from instrument channels which were not connected to the DAS.
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REVISION gq NUMBFR NUMBER Fermi 2 05000 341 OF 8
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- - 001 00 TEXT (u more nace as reearreo. use acost onet copm or Nec Form ass 9 (s1) f As a result, this event did not affect the safe operation of the plant or safety of the public. If actual conditions had existed which required the trip of HPCI and RCIC, the trips would have occurred and the safe operation of the plant and the safety of the public would have been assured.
Corrective ActionM:
The DAS instruments were placed under control to ensure that they will not be connected to plant instmmentation where the installation may be intmsive. Lessons leamed from this event and other events related to monitoring and test equipment use and control will be communicated to Plant Engineering, Technical Engineering, Maintenance, Operations, and Work Control personnel during lessons leamed training. Industry good practices regarding test control and use of monitoring and test equipment will be evaluated to identify any necessary programmatic improvements. In the interim, the temporary modification procedure will be clarified to distinguish between the use of test equipnient to take momentary readings and the actual connection of data acquisition and monitoring equipment requiring the use of temporary modification process.
Previous Similar Events
There are no previous similar events.
NFC FORM 3C6A (S-92!