05000341/LER-2008-002, Regarding Secondary Containment to Suppression Chamber Vacuum Breaker Functionality
| ML081200594 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 04/22/2008 |
| From: | Plona J DTE Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NRC-08-0030 LER 08-002-00 | |
| Download: ML081200594 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3412008002R00 - NRC Website | |
text
Joseph K, Plona Site Vice-President 6400 N. Dixie Highway, Newport, MI 48166 Tel: 734.586.5910 Fax: 734.586.4172 DTE Energy-10 CFR 50.73 April 22, 2008 NRC-08-0030 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington D C 20555-0001
Reference:
Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43 Subject: Licensee Event Report No. 2008-002, "Secondary Containment to Suppression Chamber Vacuum Breaker Functionality" Pursuant to 10 CFR 50.73(a)(2)(v)(D), Detroit Edison'is hereby submitting the enclosed Licensee Event Report (LER) No. 2008-002. This LER documents a conditioh that impacted the safety function of the reactor building to suppression chamber vacuum breaker isolation valves.
No commitments are made in this LER.
Should you have any questions or require additional information, please contact Mr. Ronald W. Gaston of my staff at (734) 586-5197.
Sincerely, cc:
NRC Project Manager NRC Resident Office Reactor Projects Chief, Branch 4, Region III Regional Administrator, Region III Supervisor, Electric Operators, Michigan Public Service Commission
NRC FORM 366' U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: No. 3150-0104 Expires 8/31/2010 (9-2007)
, the NRC may (See reverse onot conduct or sponsor, and a person is not required to respond to, the digits/characters for each block) information collection.
- 3. PAGE Fermi 2 05000341 1 OF 4
- 4. TITLE Secondary Containment to Suppression Chamber Vacuum Breaker Functionality
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NO 05000 03 12 2008 2008 002 00 04 22 2008 FACILITY NAME DOCKET NUMBER 05000
- 9. OPERATING MODE
- 11. THIS REPORT SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply) 020.2201 (b) 0 20.2203(a)(3)(i) 0 50.73(a)(2)(i)(C) 0 50.73(a)(2)(vii) 1 020.2201 (d) 0 20.2203(a)(3)(ii) 0 50.73(a)(2)(ii)(A) 0 50.73(a)(2)(viii)(A)
Q 20.2203(a)(1) 0 20.2203(a)(4) 0 50.73(a)(2)(ii)(B) 0 50.73(a)(2)(viii)(B)
El 20.2203(a)(2)(i) 0 50.36(c)(1 )(i)(A) 0 50.73(a)(2)(iii) 0 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL EJ 20.2203(a)(2)(ii) 050.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A) 0 50.73(a)(2)(x)
EJ 20.2203(a)(2)(iii) 0 50.36(c)(2) 0 50.73(a)(2)(v)(A) 0 73.71 (a)(4) 100%
0 20.2203(a)(2)(iv) 0 50.46(a)(3)(ii) 0 50.73(a)(2)(v)(B) 073.71 (a)(5) 0 20.2203(a)(2)(v) 0 50.73(a)(2)(i)(A) 0 50.73(a)(2)(v)(C) 0 OTHER 0
20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B) 50.73(a)(2)(v)(D)
Specify in abstract below or in (If more space is required, use additional copies of NRC Form 366A)
Initial Plant Conditions
Mode Reactor Power 100 percent Description of the Event During the performance of Preventive Maintenance event T272070100 for T23N010B, Division 2 Reactor Building to Torus Vacuum Breaker [BF] Isolation Valve Pressure Differential Switch [PDS],
the differential pressure switch failed calibration. On March 13, 2008 at 1445 hours0.0167 days <br />0.401 hours <br />0.00239 weeks <br />5.498225e-4 months <br />, an engineering analysis identified that the differential pressure switches (T23NO10A and B) for both Reactor Building to Torus Vacuum Breaker Isolation Valves (T2300F409 and T2300F410) would not perform their design function. This condition would have prevented the fulfillment of the safety function of structures or systems needed to mitigate the consequences of an accident. The condition impacted the function of both reactor building to suppression chamber vacuum breaker isolation valves.
Fermi 2 is a General Electric Boiling Water Reactor (BWR) with a Mark I primary containment comprised of a Drywell and Suppression Chamber. The primary containment is designed for an external pressure of 2 psi. The function of the reactor building to suppression chamber vacuum breakers is to relieve vacuum when primary containment depressurizes below reactor building pressure. The design of the reactor building to suppression chamber vacuum relief line consists of two vacuum breakers (an air operated butterfly isolation valve [ISV] and a self actuating vacuum breaker [VACB]), located in series in each of two penetrations from the reactor building to the suppression chamber airspace. The butterfly valve is actuated by differential pressure sensors which result in air operated actuators opening the isolation valve. The two vacuum breakers in series are normally closed to maintain a leak tight primary containment boundary. The butterfly valve is actuated by a differential pressure signal of 0.25 psid and the vacuum breaker opens at a maximum differential pressure of 0.5 psid. The valves and piping in either of the two penetrations are sized to provide sufficient flow to equalize the pressure between the suppression chamber and the reactor building during postulated accident events involving failure to secure drywell spray following a steam leak in the drywell and following a design basis accident (DBA).
Technical Specification (TS) 3.6.1.7, Reactor Building to Suppression Chamber Vacuum Breakers, Condition D, was entered for both reactor building to suppression chamber vacuum breaker lines with one or more vacuum breakers inoperable for opening. The Condition D action was to restore vacuum breakers in one line to an operable status within one hour. After it was determined replacement switches were available, one Reactor Building to Torus Vacuum Breaker Isolation Valve was failed to the open position, restoring operability of that line and exiting Condition D. With one vacuum breaker failed IU.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL REVISION Fermi 2 05000341 NUMBER NUMBER 3 OF 4 2008 002 00 open, TS 3.6.1.7 Condition A required closing the vacuum breaker within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. At 2115 hours0.0245 days <br />0.588 hours <br />0.0035 weeks <br />8.047575e-4 months <br /> on March 15, 2008 replacement differential pressure switches were installed and tested and TS 3.6.1.7 limiting condition for operation was exited.
Sig~nificant Safety Consequences and Implications There were no nuclear safety consequences as a result of this event since the reactor building to suppression chamber vacuum breaker isolation valves were not called upon to perform their safety function.
The reactor building to suppression chamber vacuum breaker isolation valves are in series with the reactor building to suppression chamber vacuum breakers and automatically open to establish a flowpath to the vacuum breakers for equalizing the pressure between the suppression chamber and the reactor building during postulated accident events involving failure to secure drywell spray following a steam leak in the drywell and following a design basis accident (DBA). Termination of containment spray is a procedure action, directed during execution of the Emergency Operating Procedures (EOPs).
The primary containment component with the lowest allowable external pressure is the torus shell above the suppression chamber pool water level. Failure will not result in loss of suppression pool water inventory. This failure would not prevent the Emergency Core Cooling Systems (ECCS) from performing their function of mitigating core damage; rather primary containment release mitigating capability would be compromised.
There is a small probability the reactor building to suppression chamber vacuum breakers will be required to operate. Should they be required to operate, failure of the reactor building to suppression chamber vacuum breakers would impact only the containment integrity risk metric, Large Early Release Frequency (LERF). The calculated value for the change in LERF is very small, 8.4E-8/yr, which is less than the limit of 1.OE-7/yr. This demonstrates that the nuclear safety significance is very low.
This event is being reported under 10 CFR 50.73(a)(2)(v)(D), as a condition that could have prevented the fulfillment of the safety function of a system needed to mitigate the consequences of an accident. An 8-hour non-emergency notification was made to the NRC at 18:24 EST on March 13, 2008 (EN 44062).
Cause of the Event
The condition that could have prevented fulfillment of the safety function of the reactor building to suppression chamber vacuum breakers was caused by differential pressure switch operation outside the qualified range. The differential pressure switch range was 0 to 10 inches of water column with normal suppression chamber pressure approximately 15 inches of water column on the low side of the switch.
This caused an under-range condition during normal operation resulting in a hydraulic lockup preventing the differential pressure switch from responding to a change in pressure.
1U.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL REVISION Fermi 2 05000341 NUMBER NUMBER 4 OF 4 2008 002 00 Deficiencies in the equivalent replacement process resulted in differential pressure switches being installed in the plant that could not perform their intended function during normal and accident conditions. Problems were noted in the design evaluation and testing process.
Corrective Actions
The differential pressure switches were replaced with a model having a range that encompasses the normal suppression chamber pressure.
This event is documented and evaluated in the Fermi 2 corrective action program. A root cause evaluation was performed to address this event. Reviews of internal corrective action documents and operating experience did not identify any additional cases where an equivalent replacement part was installed in the plant in an application outside of its design. This event is unique to differential pressure switches that have been installed where normal plant conditions would place the instrument in an over-range protection mode. By design, the switch operating range should bound normal plant conditions.
The deficiencies noted in the equivalent replacement process were reviewed for similar deficiencies in the engineering design change process. Similar deficiencies were not present. Additional actions are being taken to strengthen the equivalent replacement process to include a more thorough review of Operating Experience and improving the post modification testing process to prevent recurrence of this event. Corrective actions will be tracked and implemented as necessary in accordance with the Fermi 2 Corrective Action Program.
Additional Information
A. Failed Components: ITT Barton 581A-2/199 B.
Previous LERs on Similar Problems:
None I