05000529/LER-2003-004
| Docket Number | |
| Event date: | 12-08-2003 |
|---|---|
| Report date: | 05-27-2004 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(V) |
| 5292003004R00 - NRC Website | |
1. REPORTING REQUIREMENT(S):
This LER is being submitted pursuant to 10 CFR 50.73(a)(2)(i)(B), to report an LCO 3.0.4 technical specification (TS) violation that occurred when the unit entered a specified mode without the required LCO for Auxiliary Feedwater (AF) being met. LCO 3.0.4 requires that when an LCO is not met, entry into a MODE or other specified condition in the Applicability shall not be made except when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time.
LCO 3.7.5, Auxiliary Feedwater (AFW) System requires three AFW trains for Modes 1, 2 and 3. The required action for an inoperable AFW train does not allow continued operation for an unlimited period of time.
2. DESCRIPTION OF STRUCTURE(S), SYSTEM(S) AND COMPONENT(S):
The AFW System consists of one essential motor driven AF pump "B" (EIIS code: BA, P, MO), one non-essential motor driven AF pump "N" (EIIS code: BA, P, MO), and one essential steam turbine driven AF pump "A" (EIIS code: BA, P, TRB) configured into three trains. This system is used to supply feedwater to the steam generators (SG)(EIIS Code:
SG) during various plant operations. The two essential trains of the AF system automatically supply feedwater to the steam generators to remove decay heat from the Reactor Coolant System (EIIS Code: AB) upon the loss of normal feedwater supply. The essential trains of AF rely on pump discharge check valves to close to keep the output flow of the opposite train from being diverted from the SGS in the event one of the AF pumps fails to run when called upon.
The discharge check valves are 6 inch, 900 pound, Anchor Darling tilting disk check valves.
3. INITIAL PLANT CONDITIONS:
On December 8, 2003 at approximately 0540 MST, Unit 2 was in Mode 4, Hot Shutdown, at approximately 285 degrees F and 380 psia with heat-up activities in progress to return the unit to service following a refueling outage.
4. EVENT DESCRIPTION:
On December 8, 2003 at approximately 0540 MST Operations personnel determined the A train auxiliary feedwater pump discharge check valve 2PAFAV015 failed to meet the back flow acceptance criteria of surveillance test 73ST-9X138. The valve back flow was required to be limited to less than or equal to 10 gallons per minute and the tested value was 15 gpm. The valve was declared inoperable, however, no LCO was required to be entered since TS 3.7.5 only requires one motor driven AF pump to be operable in Mode 4 and the non-class pump was in operation and operable. Operations personnel took action to isolate the check valve from the B train flow path by closing the A train pump discharge manual isolation valve.
Engineering personnel subsequently determined that the test performed on the check valve was not a valid test because the check valve had not been opened with pump flow since the last successful performance of the test. Based on this determination, Operations personnel declared the discharge check valve operable. It should be noted that this test is not required to satisfy a specific TS surveillance requirement. The test was developed to verify that an AF discharge check valve has closed after pump flow has opened the valve and is required by the 1ST program to be performed during each refueling outage.
Typically, following a refueling outage, after NOP/NOT conditions have been achieved the A train AF pump is started and verified to be capable of meeting its design function. Since this evolution results in the discharge check valve being opened with pump flow, the discharge check valve would then be tested to ensure it has closed.
On December 8, at approximately 1411 MST, the unit entered Mode 3. Due to a steam leak on a SG feedwater line the unit was cooled down to make repairs achieving Mode 4 at approximately 0147 MST on December 9, and Mode 5 at approximately 0504 MST.
After the steam leak was repaired the unit was again heated up to NOP/NOT conditions achieving Mode 4 at approximately 0553 MST on December 10, and Mode 3 at approximately 1039 MST.
The A train pump was tested and successfully demonstrated full flow capability on December 11, 2004 at approximately 0146 MST. Following this testing the A train discharge check valve, AFAV015, was tested and demonstrated zero gpm backflow on December 11 at 0345 MST thus meeting the acceptance criteria of less than or equal to 10 gpm.
On April 1, 2004, during an integrated inspection exit meeting, the resident NRC inspector stated that the NRC had concluded that a technical specification violation of LCO 3.0.4 had occurred when the unit entered Mode 3 on December 8 and again on December 10, since there was insufficient evidence to conclude that the discharge check valve was operable at the time of both Mode 3 entries.
5. ASSESSMENT OF SAFETY CONSEQUENCES:
The safety significance of the inoperable discharge check valve is low. The design AF system flow rate is 750 gpm delivered to the intact steam generator(s). The safety analysis flow rate for the AF system is 650 gpm delivered to the intact steam generator(s).
An assessment of the impact of the condition described in this LER (check flow back leakage) was performed by conservatively estimating the affect the condition would have had on the actual flow that would be delivered to the steam generators. The leakage measured during the ST was used to estimate the diversion flow that would be experienced during anticipated accident conditions. The analysis, documented in calculation 02-MA-AF-0042, conservatively estimated the diversion flow through the leaking check valve at 158 gpm (Note: This calculation does not change the design basis for the AF system. It is only for the purpose of evaluating this event). When the allowance for the check valve (158 gpm) is computed, the delivered flow to the steam generators is 544 gpm. Although this value is less than the safety analysis required flow, the value is in excess of the 500 gpm delivered flow to the steam generator(s) which is the success criteria used in the PVNGS PRA. The 500 gpm conservatively assumes reactor core conditions that maximize residual heat. The events described in this report occurred and were resolved prior to unit criticality following unit shutdown for steam generator replacement and refueling. In these circumstances, potential residual heat and, therefore, AF flow requirements would be substantially lower.
There was no safety system functional failure as defined by 10 CFR 50.73(a)(2)(V).
6. CAUSE OF THE EVENT:
The cause of the condition was human performance errors made by Engineering and Operations personnel. They incorrectly determined that since the check valve test was not required to be performed the test results could be invalidated without further evaluation and justification of valve operability.
No unusual characteristics of the work location (e.g., noise, heat, poor lighting) directly contributed to this event.
7. CORRECTIVE ACTIONS:
Revise 73ST-9X138 for Engineering to evaluate any test results that fall outside of the acceptance criteria to determine if the valve is operating acceptably and to provide Operations with alternate method(s) to fully seat the valve so that an accurate seat leakage for a sealed valve can be obtained.
Revise PVNGS Surveillance Test Program, 73DP-9ZZ14, to provide additional guidance to address out of tolerance test data and when it is appropriate to invalidate a surveillance test.
8. PREVIOUS SIMILAR EVENTS:
Two events have been reported by APS in the last three years involving a TS 3.0.4 violation.
inoperable AF pump. The cause for this event was attributed to inadequate change management for a procedure change and inadequate understanding of TS 3.7.5 by the involved Operations personnel.
percent with axial shape index outside TS limits. The cause of this event was attributed to an inadequate understanding of TS 3.2.5 by the involved Operations personnel.
9.�ADDITIONAL INFORMATION:
None