05000255/LER-2007-008, Regarding Auxiliary Feedwater Pump Inoperable in Excess of Technical Specification Requirements Due to a Postulated Steam Line Break
| ML073440206 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 12/06/2007 |
| From: | Schwarz C Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 07-008-00 | |
| Download: ML073440206 (5) | |
| 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)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 2552007008R00 - NRC Website | |
text
Entergy Nuclear Operations, Inc.
Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043 Tel269 764 2000 December 6,2007 10 CFR 50.73(a)(2)(i)(B) 10 CFR 50.73(a)(2)(v)(B)
U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Palisades Nuclear Plant Docket 50-255 License No. DPR-20 Licensee Event Report 07-008, Auxiliary Feedwater Pump Inoperable in Excess of Technical Specification Requirements Due to a Postulated Steam Line Break
Dear Sir or Madam:
Licensee Event Report (LER)07-008 is enclosed. The LER describes the discovery that a postulated steam line break occurring in the turbine building could potentially render the turbine driven auxiliary feedwater pump inoperable. The occurrence is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) and 10 CFR 50.73(a)(2)(v)(B).
Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.
Christopher J. ~ c h w k Site Vice President Palisades Nuclear Plant Enclosure ( I )
CC Administrator, Region Ill, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC
ENCLOSURE I LER 07-008 Auxiliary Feedwater Pump Inoperable in Excess of Technical Specification Requirements Due to a Postulated Steam Line Break 3 Pages Follow
ITE ONE L NRC FORM 366 U S NUCLEAR REGULATORY COMMISSION (6-2004)
LICENSEE EVENT REPORT (LER)
(See reverse for requlred number of dlgitslcharacters for each block)
.INE FOR E APPROVED BY OMB NO 3150 0104 EXPIRES 6-30 2007 Estrmated, the NRC may not conduct or sponsor, and a person is not requlred to respond to, the lnformatlon collectlon FACIUTY NAME (I)
DOCKET NUMBER (2)
PAGE (3) 3 I
d 3
Auxiliary Feedwater Pump Inoperable in Excess of Technical Specification Requirements Due to a Postulated Steam Line Break LEU NUMBER (6)
OTHER FACILITIES INVOLVE0 (8)
FACILITY NAME DOCKET NUMBER DOCKET NUMBER I -
FACILITY NAME OPERATING 5
THlS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 3: (Check all that apply) (11)
MODE (9) 1 20.2201(b) 20.2203(a)(3)(it) 1 1
5 0
50.73(a)(2)(ix)(A)
I POWER 1 20.2201(d) 20.2203(a)(4)
I 1 50.73(a)(2)(iii) 50.73(a)(2)(x)
LEVEL (1 0) 20.2203(a)(I) 50.36(c)(l )(i)(A) 1 50.73(a)(2)(lv)(A) 73.71 (a)(4) 20.2203(a)(2)(i)
I 50.73(a)(2)(v)(A) 50.36(c)(I)(li)(A) 73.71 (a)(5) 20.2203(a)(2)(1i) 50.36(~)(2)
X 50.73(a)(2)(~)(B)
OTHER 1 20.2203(a)(2)(iii)
I 1 50.46(a)(3)(it) 50.73(a)(2)(v)(C)
Specify ~n Abstract below or in 20.2203(a)(2)(iv) 1 50,73(a)(2)(i)(A) 50.73(a)(2)(v)(D)
I NRC 366A 20.2X3(a)(21(vt) -
50.73{a)(2)(viii)(A)
TELEPHONE NUMBER (Include Area Code)
(
aniel G. M -
COMPLI iACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
I MAMU-RfPQRTABLE MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT FAmUliER TO EPtX SYSTEM COMPONENT FA CTURER TO EPIX SUPPLEMENTAL REPORT EXPECTED (14)
EXPECTED MONTH 1 DAY YEAR SUBMISSION I
I On December 19, 2006, it was determined that the effects of a postulated steam line break occurring in the turbine building had not been rigorously evaluated for the potential effects on safety related components within the area; specifically, auxiliary feedwater (AFW) pumps P-8A and P-8B, and their associated support components. Subsequently, a bounding engineering analysis was performed, which considered a circumferential break of a 36-inch main steam line in the turbine building. The engineering analysis determined that the resultant ambient conditions in the turbine building constituted a harsh environment under the electrical environmental qualification program, due to the combination of steam, temperature and pressure.
A subsequent evaluation of operability on October 11, 2007, determined that P-8B would have been inoperable following a steam line break in the turbine building from the harsh environment created from building pressurization. Therefore, over the previous three years, P-8B and its corresponding AFW train were inoperable for periods of time longer than allowed by Technical Specifications.
I I
This occurrence is reportable in accordance with 10 CFR 50.73(a)(Z)(i)(B) as a condition prohibited by Technical Specifications, and 10 CFR 50.73(a)(Z)(v)(B) as a condition that could have prevented the fulfillment of the safety function of a system that is needed to remove residual heat. U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION 11
EVENT DESCRIPTION
FACILITY NAME (1)
Palisades Nuclear Plant On December 19, 2006, it was determined that the effects of a postulated steam line break occurring in the turbine building had not been rigorously evaluated for the potential effects on safety related components within the area; specifically, auxiliary feedwater (AFW) pumps P-8A and P-8B [P;BA], and their associated support components. Subsequently, a bounding engineering analysis was performed, which considered a circumferential break of a 36-inch main steam line in the turbine building. The engineering analysis determined that the resultant ambient conditions in the turbine building constituted a harsh environment under the electrical environmental qualification (EEQ) program, due to the combination of steam, temperature and pressure.
05000255 On August 10, 2007, in consideration of the engineering analysis results, a past operability assessment was completed. The conclusion of the past operability assessment was that the AFW pumps and their support components would have been able to perform their design basis functions in the event of a steam line break in the turbine building, although certain components should be'considered operable but non-conforming with respect to meeting EEQ program guidelines for qualification in a harsh environment.
TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
On October 11, 2007, a re-assessment of past operability determined that several components in the turbine building that were originally considered to be operable but non-conforming should have been considered inoperable due to uncertainty that the components would function as designed in the harsh environment. Among these components were solenoid valves SV-0522B and SV-0522C [V;SB], which must function in order to open steam supply valve CV-0522B [V;SB] for turbine driven AFW pump P-8B. Since the opening of CV-0522B is not assured, the inoperable solenoid valves ultimately result in P-8B being inoperable.
P-8B would have been inoperable following a steam line break in the turbine building except during periods when the turbine building roll-up doors were open, which would mitigate pressure effects by creating a large flow path to vent steam. In general, the roll-up doors are only closed during cold weather and during severe weather conditions such as high winds or heavy precipitation.
Therefore, over the previous three years, P-8B and its corresponding AFW train were inoperable for several periods of time longer than the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed by Technical Specification (TS) 3.7.5.A. During the period P-8B was inoperable, AFW pumps P-8A and P-8C [P;BA] were periodically made inoperable for routine testing and maintenance, resulting in less than two AFW pumps being operable for periods of time in excess of TS 3.7.5.B. Additionally, during the periods P-8B was inoperable, there was one brief occasion (< 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) when P-8A and P-8C were simultaneously inoperable due to both pumps being placed in manual for testing. This condition U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION resulted in not meeting the required action and associated completion time of TS 3.7.5.C to immediately initiate action to restore one AFW train to operable status.
FACILITY NAME (1)
Palisades Nuclear Plant This occurrence is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications, and 10 CFR 50.73(a)(2)(v)(B) as a condition that could have prevented the fulfillment of the safety function of a system that is needed to remove residual 05000255 heat.
CAUSE OF THE EVENT
TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
The potential interaction between a steam line break in the turbine building and AFW system components in either the turbine building or AFW pump room was not previously analyzed. The design basis document contained a general statement that the turbine building is of sufficient size to dissipate any energy release from a postulated main steam line failure without significant pressurization or other adverse environmental effects.
11 CORRECflVE ACTIONS i
As an interim compensatory measure, the turbine building roll-up doors were controlled in the open position to mitigate the pressure effects of a steam line break in the turbine building until a permanent modification was completed. Subsequently, a modification was completed to install a blow-out panel in the turbine building, which will function to reduce the pressure gradient and prevent moisture and steam intrusion into component internals.
1 SAFETY SIGNIFICANCE
11 The event is considered to be of very low safety significance based on a combination of low initiating event frequency, and availability of AFW from at least one alternate AFW pump.
The low initiating event frequency includes consideration that the issue was precluded during approximately 6 months of the year when the turbine building was vented by its open roll-up doors.
AFW flow remained available via operable AFW pumps P-8A and/or P-8C, except for one brief period when P-8A and P-8C were inoperable due to being in manual for testing. In that instance, operators would have been directed by procedure to recover steam generator level, and could have readily started P-8A or P-8C from the control room.
PREVIOUS SIMILAR EVENTS
I None 11