05000272/LER-2002-001
| Salem Generating Station 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)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor |
| 2722002001R00 - NRC Website | |
DOCKET (2) 02 SEZmB
ENTIAL
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0 0 1 SALEM UNIT 1 05000272
PLANT AND SYSTEM IDENTIFICATION
Westinghouse — Pressurized Water Reactor Feedwater/Steam Generator Water Level Control System {JB/-) * Plant Protection System {JC/-) * * Energy Industry Identification System {EIIS} codes and component function identifier codes appear as (SS/CCC)
CONDITIONS PRIOR TO OCCURRENCE
Salem Unit 1 and 2 were in Mode 1 at 100% power prior to this event. No structures, systems or components were inoperable at the time of the occurrence that contributed to this event.
DESCRIPTION OF OCCURRENCE
On February 15, 2002, PSEG received information from Westinghouse indicating the steam generator low- low level setpoint for reactor trip {JB/-} {JC/-} and initiation of Auxiliary Feedwater was potentially non- conservative. The information provided by Westinghouse described a Steam Generator "Mid-Deck" pressure loss, which is developed as a function of steam flow rate. This Mid-deck delta-P was not considered in the existing Salem 1 and 2 instrument uncertainty calculations. The delta-P is required to be considered as a bias in the non-conservative direction, thus impacting the existing steam generator low-low level setpoints. Mid- deck delta-P information was provided by Westinghouse for both the Model "F" (Salem Unit 1) and the Model "51" Steam Generators (Salem Unit 2), specifying the Mid-deck Plate pressure loss as a function of steam flow rate.
The steam generator low-low level trip prevents loss of secondary side heat transfer capability. The low-low level trip must be operable in Modes 1 and 2. This signal is used as a primary protection signal for the design basis loss of normal feedwater, loss of offsite power and feedwater line break safety analysis, and as a backup signal for turbine trip. At the time of the Westinghouse notification, the low-low steam generator level setpoint was set at 9%, with an allowable value of 8%, for Salem Unit 1 and 2. Subsequent evaluation of the plant specific analyses for Salem 1 and 2 determined that a power reduction was required to maintain the plants within the current safety analysis limits.
Unit 1 reactor power was lowered to less than 38% and Unit 2 reactor power was lowered to 78% when the condition was found. Lower reactor power was maintained for Salem Units 1 and 2 until the setpoints for the Steam Generator low-low level reactor trips were changed. On February 16, 2002, design changes were implemented for Salem Unit 1 and 2 to raise the Steam Generator low-low level Setpoint to 14% with an allowable value of 13%. The determined analytical values for the setpoints are actually different due to the different Mid-deck delta-P values and the instrument spans for the Model F and Model 51 steam generators, but the setpoints were made the same for both units in order to provide consistency for human factors reasons.
Unit 2 returned to full power on February 17, 2002 and Unit 1 returned to full power on February 18, 2002.
DOCKET (2) 02
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0 0 SALEM UNIT 1 05000272
CAUSE OF OCCURRENCE
Mid-Deck delta-P impact on Low-low steam generator level setpoints is attributed to previously unaccounted for differential pressure created by steam flow past the mid-deck plate in the moisture separator section of the steam generator. Westinghouse indicated that this differential pressure phenomenon would cause the steam generator narrow-range to read higher than the actual water level.
PRIOR SIMILAR OCCURRENCES
A review of LERs over the past two years for Salem Unit 1, Salem Unit 2 and Hope Creek identified no similar occurrences.
SAFETY CONSEQUENCES AND IMPLICATIONS
In the event of either a Loss of Normal Feedwater (LONF) or Loss of Offsite Power to the Station Auxiliaries (LOOP), the potential existed that the low-low steam generator (SG) level reactor trip and subsequent initiation of Auxiliary Feed Water (AFW) modeled in the LONF / LOOP transients may have been delayed or defeated as a result of the SG level bias introduced by the SG mid-deck plate pressure drop. In this scenario, other diverse trip functions, such as overtemperature AT (OTAT), would be expected to trip the unit rapidly in response to the LONF / LOOP events. Any reactor trip such as OTAT would essentially eliminate the effect of the mid-deck plate pressure drop and would then allow initiation of AFW on low-low SG level such that no damage to the core or the reactor coolant system would be expected.
Since both of the safety-grade pressurizer power-operated relief valves were operable and unblocked at the time of discovery of the mid-deck plate SG level bias, there was no potential for relief of water through the pressurizer safety valves in the event of filling the pressurizer on a LONF / LOOP transient. Therefore, this event would not have propagated from a Condition II LONF / LOOP event to a Condition III Small Break LOCA (SBLOCA) event (pressurizer safety valve [PSV] stuck open due to water relief through these valves).
Even if a PSV were to fail open via water relief, the RPS and ECCS are adequately designed to protect against a SBLOCA, and the resulting SBLOCA would be bounded by the current analysis results presented in the UFSAR.
With respect to the Feed Water Line Break (FLB) transient, the nature of the transient, with blowdown via both the SG outlet nozzle and the feed water ring, is such that no appreciable mid-deck pressure drop bias will affect the low-low SG level reactor trip function. Thus, the results of a FLB transient, assuming no change to the low-low SG level trip setpoint, would have been bounded by the results presented in the UFSAR.
DOCKET (2) SALEM UNIT 1 05000272 �
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02 0 0 1 N SAFETY CONSEQUENCES AND IMPLICATIONS (contd.) Since the mid-deck plate pressure drop affect on the low-low SG level reactor trip function was not previously considered and explicitly evaluated, Salem Units 1 and 2 were conservatively classified as being in an unanalyzed condition. Subsequent evaluations determined that the trip setpoint was OPERABLE but non- conforming provided the reactor power was limited to less than 38% reactor power for Salem Unit 1 and 85% reactor power for Salem Unit 2. By reducing reactor power, steam flow through the SG is reduced. The lower steam flow reduced the pressure drop across the SG mid-deck plate. The lower pressure drop reduced the total instrument uncertainty (random uncertainty plus SG mid-deck plate level bias) to an acceptable level based upon the current Technical Specification allowable setpoint.
Thus, there were no actual safety consequences associated with this event, and this event did not present an increased risk to the health and safety of the public.
CORRECTIVE ACTIONS:
1. Salem Units 1 and 2 immediately reduced plant power levels to a value that corresponded to a point where the non-conservative Process Measurement Accuracy (PMA) term did not impact the function of the low-low level trip function.
2. Design Changes were issued and implemented to revise the Steam Generator low-low level setpoints from 9% to 14%.
3. Emergency Operating Procedure (EOP) and Emergency Response Guideline (ERG) bases were reviewed for impact on operator actions that are currently taken based on indicated steam generator level.
4. A Technical Specification change to reflect new Steam Generator low-low level setpoints will be submitted.
COMMITMENTS
The corrective actions cited in this LER are voluntary enhancements and do not constitute commitments.