05000388/LER-2005-004, Re Common Mode Failure of Inboard a and B Loop RHR Shutdown Cooling Testable Check Valves Due to Vibration-Induced Seat Damage
| ML052090373 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 07/18/2005 |
| From: | Saccone R Susquehanna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| PLA-5908 LER 05-004-00 | |
| Download: ML052090373 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) 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) 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3882005004R00 - NRC Website | |
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Robert A. Saccone PPL Susquehanna, LLC I I it Vice President - Nuclear Operations 769 Salem Boulevard e I Berwick, PA 18603 U a..
Tel. 570.542.3698 Fax 570.542.1504 rasaccone~pplweb.com l
JUL 18 2005 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop OPI-17 Washington, DC 20555 SUSQUEHANNA STEAM ELECTRIC STATION LICENSEE EVENT REPORT 50-388/2005-04-00 LICENSE NO. NPF-22 PLA-5908 Docket No. 50-388 Attached is Licensee Event Report (LER) 50-388/2005-04-00. This event was determined to be reportable in accordance with 10 CFR 50.73(a)(2)(vii).
On March 2, 2005, the containment penetration for the 'A' RHR Shutdown Cooling line failed to pressurize during as-found Local Leak Rate testing (LLRT). On March 4, 2005, during performance of the as-found LLRT for the 'B' RHR Shutdown Cooling containment penetration line, the penetration failed to pressurize. An investigation concluded that the LLRT failures were due to leakage past the inboard HV25 1F050A and B RHR Shutdown Cooling testable check valves. Both valves were disassembled for examination and were found to have damage on the body and disk seats. Analysis determined that the seat damage was caused by fretting of the valve seats due to cyclic disk motion during plant operation. The Unit 2 RHR HV251FO5OA and B testable check valves were subsequently disassembled and repaired. Satisfactory as-left LLRT results were obtained.
This event resulted in no actual adverse consequences to the health and safety of the public.
No commitments are associated with this LER.
Rob Saccone Vice President - Nuclear Operations Attachment Document Control Desk PLA-5908 cc:
Mr. S. Collins Regional Administrator U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. F. W. Jaxheimer Sr. Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 35 Berwick, PA 18603-0035 Mr. R. Osborne Allegheny Electric Cooperative P. 0. Box 1266 Harrisburg, PA 17108-1266 Mr. R. R. Janati Bureau of Radiation Protection Rachel Carson State Office Building P. 0. Box 8469 Harrisburg, PA 17105-8469
RC FORM 366 APPROVED BY OMB: NO. 31 50-0104 EXPIRES: 06/30'2007 84a04)
U.S. NUCLEAR REGULATORY COMMISSION
, the NRC digits/characters for each block) may not conduct or sponsor, and a person is not required to respond to, the rniIlrenn
- 1. FACILITY NAME Susquehanna Steam Electric Station Unit 2
- 2. DOCKET NUMBER
- 3. PAGE 05000388 1 OF5
- 4. TITLE Common Mode Failure of the Inboard 'A' and 'B' Loop RHR Shutdown Cooling Testable Check Valves due to Vibration-induced Seat Damage
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILIES INVOLVED MONTH DAY YEAR YEAR lSEQUENTIAL I REV MONTH DAY YEAR FACILITY NAME DOCKET NUMBER lNUMBER lNO._
FACILITY NAME DOCKET NUMBER 03 02 2005 2005 04 00 07 18 2005
- 9. OPERATING MODE 11.THIS REPORTIS SUBPMITED PUFRSUANTiTOTHE REOUIREMIENTS OF 10 CFR §: (Check all thatapply) 0 20.2201 (b) 0 20.2203(a)(3)(i) 0 50.73(a)(2)(i)(C)
[ 50.73(a)(2)(vii)
- 10. POWER LEVEL D 20.2201(d) 0 20.2203(a)(3)(ii) 0 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A) l/0 20.2203(a)(1)
El 20.2203(a)(4) 0 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(vlii)(B) 0 20.2203(a)(2)(i) 0 50.36(c)(1)(i)(A)
D 50.73(a)(2)(iii) 0 50.73(a)(2)(ix)(A) o 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A) a 50.73(a)(2)(iv)(A) 0 50.73(a)(2)(x) 0 20.2203(a)(2)(iii) 0 50.36(c)(2) 0 50.73(a)(2)(v)(A) 0 73.71(a)(4) 0 20.2203(a)(2)(iv)
[I 50.46(a)(3)(ii) 0 50.73(a)(2)(v)(B) 0 73.71 (a)(5) o 20.2203(a)(2)(v) 0] 50.73(a)(2)(i)(A) a 50.73(a)(2)(v)(C) 0 OTHER El 20.2203(a)(2)(vi)
El 50.73(a)(2)(i)(B)
[l 50.73(a)(2)(v)(D)
Specify in Abstract below or In (ff mot space is rqured, use adcltroaalccies ofNRC Form WM6)
PLANT CONDITIONS AT TIME OF EVENT Unit 2, Mode 5, 0% power
EVENT DESCRIPTION
On March 2, 2005, the containment penetration for the 'A' RHR Shutdown Cooling line failed to pressurize during as-found Local Leak Rate testing (LLRT). On March 4, 2005, during performance of the as-found LLRT for the 'B' RHR Shutdown Cooling containment penetration line, the penetration failed to pressurize. An investigation concluded that the LLRT failures were due to leakage past the inboard HV251 F050A and B RHR Shutdown Cooling testable check valves. Both valves were disassembled for examination and were found to have significant damage on the body and disk seats.
PPL conducted an investigation, including analysis performed by Structural Integrity and Kalsi Engineering, to determine the cause of the valve damage. The investigation was completed on May 31, 2005, and concluded that the damage was caused by fretting of the valve seats due to cyclic disk motion during plant operation.
This LER is being submitted in accordance with 10 CFR 50.73(a)(2)(vii) for any event where a single cause or condition resulted in at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system.
CAUSE OF THE EVENT
The cause of the body seat and disk seat damage was due to relative motion between the disk and the seat, which resulted from a combination of the following circumstances:
Leakage through the Unit 2 inboard RHR HV251 F050A and B testable check valves and/or leakage past the F1 22 bypass valve enables the section of pipe between the check valves and the HV251 F01 5A and B outboard isolation valves to pressurize, eventually to a point of equilibrium with pressure on the reactor side. Upon reaching the point of equilibrium, the only force keeping the swing check valve disk closed is gravity. The valve disk is then free to move relative to the seat as a result of structural vibration and/or hydraulic Recirculation System pump pressure pulsations.
Extended operation at high core flows accelerates the damage mechanism. Since there is an incubation period associated with fretting failure, the damage was only observed after cumulative operation at high Recirculation System pump speeds associated with high core flow.
Side-to-side disk clearance greater than vendor recommended specification on the HV251 F050B valve aggravated the relative motion of the valve disk.
NRC FORM 366 (6-2004)
(If ore space is quired, use ad*onalcopiesof NRI ANALYSIS / SAFETY SIGNIFICANCE The Unit 1 and 2 inboard RHR testable check valves HV151(251)FO50A and B are 24-inch Atwood-Morrill swing check valves with cast stainless steel bodies and disks, and stellite hard facing on the body and disk seats. The testable check valves are typically in standby service. The valves are stroked approximately eight times per operating cycle for surveillance testing and are opened with system flow for shutdown cooling for approximately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> per cycle.
X-130(B) Cantairnent Penetratin,
Loop 'A' Testable Check Valve FI4s1FD5oA(B)
H" S1F15OB)
RHR OtAboard Isolation Valve MHS1F122APf Bypass Valve Figure 1 The LLRT is performed such that it tests the leakage through both the RHR HV151(251)FO50A (or B) testable check valves and the HV1 51(251)F1 22A (or B) bypass valves (Figure 1). A review of previous Unit 2 LLRT results show that the Unit 2 RHR HV251 F050A testable check valve successfully passed its LLRT or the 1 000#
leak test during the last two refueling outages. However, prior to this recent LLRT failure, the 'B' RHR testable check valve failed its LLRT during the last two refueling outages. One of the failures was attributed to the RHR 50B bypass valve F122B. No seat damage was noted, but due to radiation dose considerations, the 'B' RHR testable check valve disk was replaced and the valve rebuilt. The second failure was attributed to the 'B' RHR check valve. Although no seat damage was found, the valve seat was lightly lapped and the disk was replaced.
A review of the Unit 1 LLRT history identified that the RHR testable check valves (i.e., HV151 F050A and B) were tested during the last three refueling outages and the results were acceptable. The valves have no prior record of seat damage. The Unit 1 RHR injection check valves are not expected to be subject to the same NRC FORM 366 (6-2004)
(if more space Is required, use atonal copes of NRC Fon 364)
- Vibration monitoring instrumentation was installed inside the Unit 2 drywell on selected points in the RHR and Reactor Recirculation systems to allow for vibration monitoring during the fuel cycle.
- Core flow restrictions have been implemented for the current Unit 2 fuel cycle to prevent damage to the Unit 2 RHR HV251 F050A and B inboard testable check valves.
Planned Actions
- Procedure TP-264-034, Reactor Recirculation/RHR Injection Loop Hydraulic Response evaluation will be performed allow stroking of the Unit 2 HV251 F01 5A and B outboard isolation valves at power and monitoring of the vibration effects.
- Evaluate a possible modification solution to upgrade the Unit 1 RHR HV151 F015A and B outboard isolation valves and/or the Unit 2 RHR HV251 F050A and B testable check valves to solve the issue of swing check valve disk positive closure.
ADDITIONAL INFORMATION
Past Similar Events:
None NRC FORM 366 (6-2004)