05000327/LER-2002-001
| Sequoyah Nuclear Plant (Sqn) Unit 1 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor |
| 3272002001R00 - NRC Website | |
I. PLANT CONDITION(S)
Units 1 and 2 were in power operation at approximately 100 percent reactor power.
II. DESCRIPTION OF EVENT
A. Event:
On February 15, 2002, Sequoyah performed an assessment of the narrow range steam generator level measurement instrument channels (EIIS Code AB) based on a Nuclear Safety Advisory Letter from Westinghouse Electric Corporation. This assessment determined that the demonstrated accuracy calculation for low-low level trip setpoint narrow range span did not account for the measurement bias associated with the differential pressure (dP) created by the steam flow past the mid-deck plate in the moisture separator section of the steam generators.
This dP phenomena could cause the steam generator narrow range level channels to read higher than actual water level at high steam flows. This could cause the low steam generator level trip setpoint to be nonconservative.
Sequoyah has an environmental allowance modifier feature that changes the low-low level steam generator setpoint.
With this feature actuated the low-low setpoints are conservative and allow continued operation. As a conservative measure, the environmental allowance modifier will remain actuated until the steam generator setpoints are evaluated and modified, if applicable.
B. Inoperable Structures, Components, or Systems that Contributed to the Event:
None.
C. Dates and Approximate Times of Major Occurrences:
February 9, 2002 February 15, 2002 February 15, 2002 February 15, 2002 at 1811 Eastern Standard Time (EST) February 15, 2002 at 1814 EST.
February 16, 2002 Diablo Canyon issues an operating experience item resulting from the failure of the reactor to trip on a low-low steam generator level.
Westinghouse issued Nuclear Safety Advisory Letter (NASL) 02-3 concerning the steam generator water level setpoint analysis.
Sequoyah determined that the demonstrated accuracy calculation did not account for the measurement bias associated with the differential pressure created by the steam flow past the mid-deck plate in the moisture separator section of the steam generators.
As a conservative measure, Operations personnel entered limiting condition for operation (LCO) 3.0.3 for potentially nonconservative steam generator low-low level setpoint on Unit 2. The environmental allowance modifier was determined to be actuated on Unit 1 since Unit 1 was venting at the time.
The environmental allowance modifier was actuated on Unit 2 and Operations personnel exited LCO 3.0.3.
Plant procedures were revised to keep the environmental allowance modifier actuated, until the steam generator setpoints are evaluated and modified, if applicable.
D. Other Systems or Secondary Functions Affected:
None.
E. Method of Discovery:
Engineering review of the Westinghouse Nuclear Safety Advisory Letter on steam generator level concluded that insufficient margin existed in the present setpoint analysis to overcome the error in the Westinghouse calculation.
F. Operator Actions:
Operations personnel entered LCO 3.0.3 for Unit 2 and then actuated environmental allowance modifier.
G. Safety System Responses:
Not applicable - no safety system response was required.
III. CAUSE OF THE EVENT
A. Immediate Cause:
The immediate cause of this condition was the error in the Westinghouse low-low level setpoint calculation potentially resulting in the Sequoyah setpoint being nonconservative.
B. Root Cause:
The root cause of the condition was the failure of Westinghouse personnel to account for the measurement bias associated with the differential pressure created by the steam flow past the mid-deck plate in the moisture separator section of the steam generators, when establishing the low-low steam generator level setpoint calculation.
C. Contributing Factor:
None.
IV. ANALYSIS OF THE EVENT
The Sequoyah FSAR accident and transient analyses credit reactor trip on low SG level for the loss of normal feedwater (LONF), loss of off-site power (LOOP), and feedwater line break transients. Since the dP phenomena does not exist for the feedwater line break transient, this analysis is not affected.
In the unlikely event of a loss of main feedwater or a loss of offsite power, the potential exists that the required reactor trip on low steam generator level may be delayed or not be received. In this scenario, alternate reactor trip signals actuating on over-temperature differential temperature, high pressurizer pressure or high pressurizer level would most likely be generated prior to the loss of nucleate boiling and actuation of the pressurizer safety relief valves. However, these reactor trip signals would arrive at a different point in the transient than currently analyzed and the net effect of the delay on the UFSAR Chapter 15 accident analysis results is unknown. Thus, a condition exists in which a credited safety function could have failed to operate.
V. ASSESSMENT OF SAFETY CONSEQUENCES
The function of the reactor protection circuits associated with low-low steam generator water level and low feedwater flow is to preserve the steam generator heat sink for removal of long term residual heat. Reactor trips on RCS temperature and pressurizer pressure will trip the unit before there is any damage to the core or the reactor coolant system. For the LONF or LOOP event, the consequences are bounded by a small break loss of coolant accident (LOCA) and therefore, the effect of the mid-deck pressure loss on this event does not represent a substantial safety hazard.
Concerning component integrity, each steam generator is analyzed for one occurrence of steam generator dryout.
For the ATWS Mitigation System Actuation Circuitry (AMSAC) system, an increase in steam generator level uncertainty could impact the AMSAC actuation setpoint and consequently, the operation of AMSAC.
The AMSAC actuation setpoint is not directly assumed in any safety analysis. AMSAC is set to operate upon reaching a low level setting of 8 percent of SG Narrow Range. The additional error added to the loop error results in a total instrument channel uncertainty of 7.6 percent of span. Even for the worst case, the AMSAC would still actuate before reaching the safety limit of 0 percent of span. The effect of this additional bias would be a slight delay in AMSAC actuation.
Steam line breaks (SLBs) outside containment producing mass and energy releases assume a reactor trip on a low-low steam generator water level as a convenience trip if no other trip signals are received. The effect of any decrease in the water level reactor trip setpoint (if greater than 0 percent of the narrow range span, and not bounded by the increase in the calculated uncertainty) is bounded by the very conservative assumptions in the analysis related to main feedwater flow and steam generator tube uncovery. The calculated SLB mass and energy releases outside containment remain conservative.
Additionally, Engineering personnel reviewed previous plant trips on low-low steam generator level, and no observable bias was found to exist.
VI. CORRECTIVE ACTIONS
A. Immediate Corrective Actions:
As a conservative measure, the environmental allowance modifiers were actuated. Plant procedures were revised to keep the environmental allowance modifier actuated, until the steam generator setpoints are modified, if applicable.
B. Corrective Actions to Prevent Recurrence:
TVA will modify the steam generator setpoints, if applicable, to account for the identified condition.
VII. ADDITIONAL INFORMATION
A. Failed Components:
None B. Previous LERs on Similar Events:
A review of previous reportable events for the past three years did not identify any previous events.
C. Additional Information:
None D. Safety System Functional Failure:
This condition is considered to be a safety system functional failure in accordance with NEI 99-02, because of the low-low steam generator setpoints being nonconservative.
VIII. COMMITMENTS
None.