05000266/LER-2007-008
| Point Beach Nuclear Plant Unit 1 | |
| Event date: | 10-19-2007 |
|---|---|
| Report date: | 12-17-2007 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| Initial Reporting | |
| ENS 43750 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident |
| 2662007008R00 - NRC Website | |
On October 19, 2007, as part of ongoing corrective actions associated with pressure and temperature limits report (PTLR) issues, an issue with the Point Beach Nuclear Plant (PBNP) Unit 1 and Unit 2 low temperature overpressure protection systems (LTOP) actuation was identified. A recent calculation showed that the maximum actuation setpoint may be non-conservative. Changes to the system parameters inputs were incorporated in the new calculations, which included:
1)The mass input from the safety injection pumps was significantly higher (25%) based on the use of a PBNP-specific system flow model versus the previously used generic Westinghouse flow model, 2) The previous setpoint calculation did not consider instrument delay times in the opening of the power-operated relief valves (PORV), and 3) Instrument uncertainties had changed.
On October 25, 2007, at 1930 CDT, PBNP Unit 1 and Unit 2 LTOP systems were declared inoperable based on the new calculation results. The new calculation resulted in a lower maximum LTOP setpoint of 5420 psig vs. the previous 5500psig. Additional operating restrictions were required to utilize 5420 psig as the maximum LTOP setpoint.
These operating restrictions are that only one reactor coolant pump may be in operation 5180°F and only two charging pumps may be in operation whenever the LTOP is enabled. These changes were incorporated into the operating procedures. The procedure changes provide the guidance required to ensure that the current LTOP setpoints remain conservative. Based on the issuance of the revised procedures with the recommended restrictions, operability of the LTOP system was restored for Unit 1 and Unit 2 on October 26, 2007, at 1751 CDT.
This condition was reported via an 8-hour non-emergency report, EN 43750, on October 25, 2007, pursuant to 10 CFR 50.72(b)(3)(v)(D), as a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident Upon further review, it was recognized that this condition should have been reported pursuant to 10 CFR 50.72(b)(3)(ii)(A), as a condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. Accordingly, this 60-day written report is submitted pursuant to 10 CFR 50.73(a)(2)(ii)(A).
Initial Conditions:
Units 1 and 2 were in MODE 1 at 100% power.
gIRC FORM 366A� U.S. NUCLEAR REGULATORY COMMISSION (9-2007) I� SEQUENTIAL I� REV Description of Occurrence:
A new calculation was performed by Westinghouse to update the PBNP PTLR with new pressure-temperature (PT) curves and LTOP setpoints that would be applicable to the end of the Renewed Licenses for Units 1 and 2. This new calculation took into account the additional fluence for the period of extended operation. The new calculation used a different methodology, Master Curve, for adjusted reference temperature (ART) values. It is recognized that the results of this calculation could not be used until the Master Curve methodology had been submitted to the Commission and was approved for use at PBNP.
During the operability review of the LTOP systems, another difference between the current and proposed LTOP setpoint calculations was identified. The previous calculation of the maximum allowable reactor pressure at low temperatures used the limiting material properties in the beltline of the reactor vessel. The new calculation uses the 10 CFR 50, Appendix G requirement that the vessel pressure must be less than or equal to 20% of the preservice system hydrostatic test pressure at low temperatures.
The operating restrictions identified by the new calculation are primarily due to the above error. Since the Master Curve methodology was not the source for the operating restrictions, the potential impact on the existing LTOP setpoint was evaluated. The recommended LTOP operating restrictions were then added to plant procedures, and the lower maximum LTOP setpoint was incorporated in the current PT curves.
Apparent Cause:
The cause of this event was determined to be personnel error. The previous calculation incorrectly considered the reactor vessel beltline material to be limiting for LTOP setpoints instead of the reactor vessel The previous calculation also contained the following non-conservative inputs/assumptions: 1) Used generic Westinghouse flow model; 2) Did not consider instrument delay times in the opening of the power-operated relief valves (PORVs); and 3) Instrument uncertainties that had changed.
These inputs/assumptions were corrected in the new calculation. The restrictions were implemented in the site operating procedures because the resulting operating restrictions contained in this calculation were not dependent upon the approval and implementation of the proposed Master Curve methodology.
.., The LTOP maximum setpoint was reduced from 5500 psig to 5420 psig. The actual setpoint of the PORVs remains at 415 psig. Two operational restrictions were placed in the operating procedures. When reactor coolant system temperature is 5180° F, only one reactor coolant pump is allowed to be operating; and when LTOP is enabled, only two charging pumps are allowed to be operating.
The change to the maximum pressure setpoint and the operational changes meet the operational requirements in the calculation and ensure that the LTOP setpoint is conservative.
Safety Assessment:
The overall safety significance of this event is low. The primary difference between the two calculations (91.5 psig of a total 97.7 psig change) was a result of the 10 CFR 50 Appendix G flange requirements vs.
use of the beltline as the limiting material.
Per WCAP-14040-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision 4, Section 2.9 states:
"10 CFR Part 50, Appendix G contains the requirements for the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTNDT by at least 120°F for normal operation when the pressure exceeds 20 percent of the pre-service hydrostatic test pressure, 3106 psig. This pressure is 621 psig for a typical Westinghouse reactor vessel design.
"This requirement was originally based on concerns about the fracture margin in the closure flange region.
During the boltup process, stresses in this region typically reach over 70 percent of the steady-state stress, without being at steady-state temperature. The margin of 120°F and the pressure limitation of 20 percent of hydrostatic test pressure were developed using the Kia fracture toughness, in the mid 1970s.
"Improved knowledge of fracture toughness and other issues which affect the integrity of the reactor vessel have led to the recent change to allow the use of K1 in the development of PT curves, as contained in Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1.
The discussion given in WCAP-15315, "Reactor Vessel Closure HeadNessel Flange Requirements Evaluation for Operating PWR and BWR Plants," concluded that the integrity of the closure head/vessel flange region is not a concern for any of the operating plants using the K1 toughness. Furthermore, there are no known mechanisms of degradation for this region, other than fatigue. The calculated design fatigue usage for this region is less than 0.1, so it may be concluded that flaws are unlikely to initiate in this region.
3 PBNP uses K10 in the development of the PT curves, as contained in ASME Code Case N-641, "Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System Requirements for Section XI, Division 1." Use of ASME Code Case N-641 at PBNP was approved by an NRC Safety Evaluation dated October 6, 2000.
An evaluation is being performed to determine whether the non-conservative maximum setpoint value posed a concern relating to past operability. A supplemental report may be submitted pending the results of this evaluation.
Additional Information:
None
Previous Occurrences:
The following licensee event reports have been submitted during the past three years associated with analysis issues:
and Non-Conservative Degraded Voltage Time Delay Relay Setting Technical Specification