05000321/LER-2004-004

LER-2004-004, Momentary False Low Reactor Water Level Signal Results in Actuations of Safety Systems

Edwin I. Hatch Nuclear Plant - Unit 1
Event date:	03-09-2004
Report date:	05-05-2004
Reporting criterion:	10 CFR 50.73(a)(2)(iv)(A), System Actuation
3212004004R00 - NRC Website

U.S. NUCLEAR REGULATORY COMMISSION FACILITY NAME (1)

DOCKET

05000-321 LER NUMBER (6) 2004 -- 004 --

PLANT AND SYSTEM IDENTIFICATION

General Electric - Boiling Water Reactor Energy Industry Identification System codes appear in the text as (EIIS Code XX).

DESCRIPTION OF EVENT

On 03/09/2004 at 0340 EST, Unit 1 was in the Cold Shutdown mode with reactor coolant temperature at

• approximately 189 degrees Fahrenheit to perform the vessel leakage test. Personnel had reduced vessel pressure from 869 psig to 129 psig in order to place the "A" loop of the Residual Heat Removal (RHR, EIIS Code BO) system into the shutdown cooling mode to facilitate repairs to a reactor vessel head vent line flange discovered leaking during the vessel leakage test. At 0340 EST, as Operations personnel were opening shutdown cooling suction valve 1E1 1 F009, reactor pressure decreased from 129 psig to 16 psig within six seconds. Due to this pressure reduction, several reactor water level transmitters sensed a sudden decrease in water level from off-scale high (greater than 60 inches above instrument zero) to 107.7 inches below instrument zero, resulting in actuations of several safety systems as designed:

• Emergency Diesel Generators (EIIS Code EK) 1A, 1B, and 1C started.

• Core Spray (EIIS Code BM) system pumps lA and 1B started and injected until manually shut down by Operations personnel.

• RHR pumps 1B and 1D started (RHR pumps 1A and 1C did not start because their suction source logic was not satisfied with valve 1E1 1F009 not open fully; RHR pumps 1B and 1D did not inject because injection by the Core Spray pumps caused reactor pressure to exceed the discharge head of the RHR pumps before they could inject).

• Valves 1P41F310A through D (EIIS Code BS) closed.

• The Main Control Room Environmental Control System (EIIS Code VI) swapped to the pressurization mode.

• A full Group 1 Primary Containment Isolation System (EIIS Code JM) signal was received and the inboard Main Steam Line Isolation Valves closed (the outboard valves already were closed).

• Reactor Water Cleanup System (EIIS Code CE) outboard isolation valve 1G31F004 closed.

• The lA Control Rod Drive (EIIS Code AA) system pump tripped.

• Valves 1B31F031A and B closed causing the lA and 1B Reactor Recirculation (EIIS Code AD) system pumps to trip.

• Shutdown cooling isolation valves 1E11F008 and 1E11F009 closed.

• The lA Residual Heat Removal Service Water (EIIS Code BS) system pump tripped.

• All four trains of the Unit 1 and Unit 2 Standby Gas Treatment (EIIS Code BH) systems started.

• An initiation signal for the Reactor Core Isolation Cooling (EIIS Code BN) system was received (the High Pressure Coolant Injection (EIIS Code BJ) logic system was tagged out at the time of the event; otherwise, it likewise would have received an initiation signal).

U.S. NUCLEAR REGULATORY COMMISSION �

DOCKET

Edwin I. Hatch Nuclear Plant - Unit 1 � 05000-321 LER NUMBER (6)

• YEAR � NUMBER No scram signal was received and Reactor Water Cleanup System inboard isolation valve 1G31F001 did not close. The lack of these actuations was expected as the water level transmitters that provide signals to the respective logic systems did not sense a low reactor water level condition.

The water level transmitters that indicated a drop in sensed reactor water level returned to their pre-event readings of off-scale high within three seconds of reaching their lowest indicated water level. Operations personnel determined correctly that the indications of a decrease in reactor water level were false and terminated injection by the lA Core Spray system pump within 36 seconds and the 1B pump within 12 seconds. Reactor pressure increased to approximately 250 psig in 12 seconds in response to injection by the two Core Spray system pumps; however, Operations personnel terminated the pressure increase when they secured both pumps. No pressure-temperature limits contained in Unit 1 Technical Specification Limiting Condition for Operation 3.4.9 were exceeded during this transient. More specifically, reactor coolant temperature (and, therefore, vessel metal temperature) remained at approximately 189 degrees Fahrenheit throughout the transient; thus, all combinations of pressure and temperature remained in the allowable regions of Figures 3.4.9-1 and 3.4.9-2 of the Unit 1 Technical Specifications.

Operations personnel systematically reset the actuation signals and returned the affected systems and components to their normal standby status by 0453 EST.

CAUSE OF EVENT

This event was caused by personnel error and procedure omissions. Personnel assembling the head vent line flange failed to check the flange gap or verify proper fit-up prior to torquing; therefore, adequate crush was not obtained on the flange gasket. Procedure 52GM-MME-004-1, "Reactor Vessel Reassembly," did not reinforce proper work practices in that it did not require personnel specifically to verify flange gap or proper fit-up. When personnel increased reactor pressure to perform the vessel leakage test, the improperly assembled flange leaked. The location of the flange leak was such that the vessel became water-solid during the vessel leakage test. Opening valve 1E11F009 with the vessel water-solid caused a pressure transient that led to a momentary false indication of low water level. Procedure 421T-TET-006-1, "ISI Pressure Test of the Class 1 System and Recirc. Pump(s) Runback Test," did not contain instructions for lowering reactor pressure and placing shutdown cooling into service with the vessel water-solid.

Prior to this event, plant personnel had increased reactor pressure for the purpose of conducting a scheduled vessel leakage test. The pressure increase was terminated at 0232 EST, when reactor pressure was 869 psig, to investigate a leak in the drywell, indicated by an increasing floor drain leakage rate. The leak, which personnel later determined to be on a flange of the reactor vessel head vent line, resulted in the expulsion of the air bubble normally present in the top of the reactor vessel during the pressure test and the vessel becoming water-solid.

Operations personnel, at the direction of site management, lowered reactor pressure to place the shutdown cooling mode of the RHR system into service so the leak could be repaired. During the time personnel FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3) lowered reactor pressure, the leak through the head vent line flange continued. When reactor pressure had been reduced to between 135 psig and 129 psig, Operations personnel opened outboard suction valve 1E11F008 in preparation for placing shutdown cooling into service. Upon reaching 129 psig, Operations personnel opened inboard suction valve 1E11F009.

As valve 1E11F009 was opening, reactor pressure dropped unexpectedly from 129 psig to 16 psig within approximately six seconds. Sensed level on at least four water level transmitters, all connected to common reference and variable legs on Condensing Chamber 1B21D003B, also dropped unexpectedly and suddenly from off-scale high (greater than 60 inches above instrument zero) to 107.7 inches below instrument zero as indicated on the Safety Parameter Display System (EIIS Code IQ). (A recorder in the Main Control Room showed that sensed water level dropped to approximately 125 inches below instrument zero.) An analysis of this event revealed that the change in sensed water level occurred only on water level transmitters connected to common reference and variable legs on Condensing Chamber 1B21D003B.

Actuations that did not occur, specifically, the lack of a scram signal and the failure of valve 1G31F001 to close, as well as Main Control Room indications and Safety Parameter Display System data, showed that no water level transient was sensed by any water level transmitter connected to common reference and variable legs on Condensing Chambers 1B21D003A, 1B21D004A, or 1B21D004B. The variable leg nozzle (tap) for Condensing Chamber 1B21D003B was the only variable leg nozzle near the suction piping for shutdown cooling.

The sensed reactor water level transient was not indicative of an actual water level transient, that is, no actual change in reactor water level occurred. Moreover, no equipment or component failure existed to explain the response of the water level transmitters in question. Instead, the water level transmitter response was the result of a pressure gradient momentarily created near the variable leg nozzle of Condensing Chamber 1B21D003B when valve 1E11F009 was opened. When personnel opened this valve, the reactor vessel was exposed to a lower pressure and perhaps partially voided RHR system volume, causing reactor pressure to decrease suddenly, which is an expected occurrence for a pressurized water-solid system. The sudden decrease in pressure, at a rate of over eighteen psi per second, momentarily created a localized lower pressure area near the variable leg nozzle on Condensing Chamber 1B21D003B. This localized low pressure area, in turn, created a momentary differential pressure between the reference and variable legs as sensed at the water level transmitters connected to these two legs. The level transmitters sensed the differential pressure as a decrease in reactor water level and generated the appropriate low water level signal per their design.

The pressure gradient dissipated before it could affect variable legs located farther from its source, that is, from the shutdown cooling suction piping. Consequently, the water level transmitters whose variable leg nozzles were located higher than the variable leg nozzle on Condensing Chamber 1B21D003B or on the opposite side of the vessel did not sense a lower pressure than their respective reference legs. No differential pressure between the reference and variable legs resulted and no sensed decrease in water level occurred.

U.S. NUCLEAR REGULATORY COMMISSION FACILITY NAME (1) � DOCKET Edwin I. Hatch Nuclear Plant - Unit 1 � 05000-321 LER NUMBER (6) 2004 -- 004 -- 0 Procedure 42IT-TET-006-1 did not contain sufficient cautions and instructions to prevent the creation of the momentary pressure gradient. More specifically, the procedure did not contain cautions and instructions necessary to recover from a situation in which the vessel becomes water-solid. Consequently, Operations personnel did not have instructions to enable them to recognize the vessel was water-solid or to lower reactor vessel pressure and place shutdown cooling into service after the vessel had become water-solid. This resulted in a significant and unexpected pressure transient that, in turn, led to a momentary false indication of low reactor water level.

REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT

This report is required by 10 CFR 50.73(a)(2)(iv)(A) because of the unplanned actuation of reportable systems. Specifically, several safety systems actuated upon receipt of a false low reactor water level signal caused by some water level transmitters responding to a sudden drop in reactor vessel pressure. All systems functioned as designed given the level signal sent to their respective logic systems. Systems and components that did not actuate either receive logic signals from unaffected water level transmitters or were removed properly from service at the time of the event.

A Level 1 low reactor water level signal is generated when sensed water level decreases to 101 inches below instrument zero (approximately 58 inches above the top of the active fuel). This signal indicates the possibility of a significant loss of reactor coolant, such as would be expected in a design basis loss-of-coolant accident. Therefore, multiple safety systems actuate automatically upon receipt of a Level 1 reactor water level signal to ensure sufficient water is available to maintain adequate core cooling and to limit the further loss of coolant inventory and the release of any radioactive material. For example, the Emergency Diesel Generators start automatically to ensure safety systems have the necessary power in the event of a coincident loss of offsite power; Core Spray and RHR system pumps start and injection valves open automatically to provide the necessary water to recover and maintain coolant inventory; primary and secondary containment isolation valves close automatically to limit the loss of coolant inventory and the release of any radioactive material.

In this event, all components and systems functioned as designed and intended given the false low reactor water level signal generated. Systems and components that did not actuate receive logic signals from unaffected water level transmitters or, in the cases of the 1A and 1C RHR system pumps and the High Pressure Cooling Injection system, were removed properly from service at the time of the event. Because the sensed low water level signal was not indicative of an actual water level transient and all safety systems would have performed their intended safety function had actual water level decreased, this event had no adverse impact on the public health or safety.

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

CORRECTIVE ACTIONS

Personnel replaced the head vent line flange gasket on 03/09/2004 and verified the flange was not leaking during a vessel leakage test completed successfully on 03/10/2004 at 0331 EST.

Procedures 42IT-TET-006-1 and 42IT-TET-006-2, "ISI Pressure Test of the Class 1 System and Recirc.

Pump(s) Runback Test," will be revised prior to their next respective performance. These revisions will provide information to help personnel determine whether the vessel is or may be water-solid. They also will include specific cautions, limitations, and instructions to recover from a suspected or actual water-solid condition.

Procedures 52GM-MME-004-1 and 52GM-MME-004-2, "Reactor Vessel Reassembly," will be revised prior to their next respective use to require that proper flange gap and fit-up be verified during flange assembly.

ADDITIONAL INFORMATION

No systems other than those previously described in this report were affected by this event.

This LER does not contain any permanent licensing commitments.

There have been no previous similar events reported in the past two years in which a false low reactor water level signal caused unplanned safety system actuations.