05000219/LER-2005-005, Re Technical Specification Violation Due to Main Steam Safety Valves Setpoints Discovered Out of Tolerance
| ML053550364 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 12/12/2005 |
| From: | Swenson C AmerGen Energy Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2130-05-20240 LER 05-005-00 | |
| Download: ML053550364 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2192005005R00 - NRC Website | |
text
AmerGen SM 45 AmerGen Energy Company www.exeloncorp.com An Exelon Company Oyster Creek US Route 9 South, P.O. Box 388 Forked River, NJ 08731-0388 10 CFR 50.73 December 12, 2005 2130-05-20240 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 - 0001 Oyster Creek Generating Station Facility Operating License No. DPR-16 NRC Docket No. 50-219
Subject:
Licensee Event Report 2005-005-00, Technical Specification Violation Due to Main Steam Safety Valves Setpoints Discovered Out of Tolerance Enclosed is Licensee Event Report 2005-005-00, Technical Specification Violation Due to Main Steam Safety Valves Setpoints Discovered Out of Tolerance. This event did not affect the health and safety of the public or plant personnel. This event did not result in a safety system functional failure. There are no new regulatory commitments made in this LER submittal.
If any further information or assistance is needed, please contact Dave Fawcett at 609-971-4284.
Sincepy-C. N. Swenson Vice President, Oyster Creek Generating Station CNS/DIF
Enclosure:
NRC Form 366, LER 2005-005-00 cc:
S. J. Collins, Administrator, USNRC Region I P. S. Tam, USNRC Senior Project Manager, Oyster Creek M. S. Ferdas, USNRC Senior Resident Inspector, Oyster Creek File No. 05054
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06i30O07 (6-200*
, the NRC may not conduct or digits/characters for each block) sponsor, and a person is not required to respond to, the Information collection.
- 3. PAGE
.Oyster:Ceek,,Unit I, 0500 19 1 OF
- 4. TITLE
.Technical Specification i o du tM Salves S
insis r
Out o fr
- 5. EVENT DATE
- 6. LER NUMBER l
- 7. REPORT DATE
. OTHER FACILITIES INVOLVED I SEQUENTIAL REV I FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO MONTH DAY YEAR 05s00 1
0 200 005 OOI 12 12 FACILrrY NAME DOCKEr NUMBER 1
105000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANTTO THE REQUIREMENTS OF 10 CFR 5: (Check all that apply) 20.2201(b) 20.2203(a)(3)(i) 50.73(a)(2)(i)(C) 50.73(a)(2)(vii) 20.2201(d) 20.2203(a)(3)Oi) 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A) 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B) 20.2203(a)(2)(i) 50.36(c)(1)(i)(A) 50.73(a)(2)(iii) 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL 20.2203(a)(2)(ii) 50.36(c)(1)(il)(A) 50.73(a)(2)(iv)(A) 50.73(a)(2)(x) 20.2203(a)(2)(iii) 50.36(c)(2) 50.73(a)(2)(v)(A) 73.71 (a)(4) 100 20.2203(a)(2)(iv) 50.46(a)(3)(1i) 50.73(a)(2)(v)(B) 73.71(a)(5) 20.2203(a)(2)(v) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(C)
OTHER 20.2203(a)(2)(vi)
X 50.73(a)(2)(i)(B) 50-73(a)(2)(v)(D)
Spec inyk Abstract below
_ _ _or__in N RC Form 368A
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME TELEPHONE NUMBER (krkde Area Code) 1Jim CorreaEngineering (609)
ONE
.'t Ah age
' 97 0 8R I--- YES (Ifyes, complete EXPECTED SUBMISSION DATE)
ABSTRACT (unda 100sa.., apprommat a-scepen Iws Oyster Creek Generating Station was in RUN at 100% power on 10/13/05 when a condition was discovered during routine laboratory as-found testing for Safety Valves (SVs) (EIIC: RV) removed during the 1 R20 refueling Outage in November 2004. There were no structures, systems or components out of service that contributed to this event.
In accordance with ASME Boiler & Pressure Vessel Code Criteria the nine SVs removed in refueling outage 1 R20 were sent for as-found setpoint testing within the one-year time frame. Based on information received from the laboratory performing SV as-found testing, Site Engineering personnel determined that SV setpoint deficiencies existed with three SVs that were installed during the 1 R1 9 refueling outage.
Three (3) of the nine (9) valves exceeded the setpoint tolerance of +1-1% (+/-12 psig) as specified in the Technical Specifications paragraph 2.3F.
All three SVs were within the ASME Code allowable +/-3% tolerance for as found values and there were no actual safety consequences associated with this event.
NRC FORM 366 (6-2004)
(If more space is required, use additional copies of (If more space is required, use additional copies of NRC Formn 366A)
Analysis of Event (cont'd)
An evaluation of the condition with regard to the Overpressure Protection Analysis does not have to be performed since the valves would have limited overpressure to below 110% (1375 psig) of design pressure (1250 psig). The average setpoint tolerance of all nine valves was -0.69% (-8.3 psig) which is within the
+/-1% limit. The Basis of Technical Specification 4.3E states that "with all safety valves set 12 psig higher the safety limit of 1375 psig is not exceeded".
This event is not considered risk significant. The applicable transient was bounded by previous analysis results therefore the safety limit would not have been exceeded.
This event is reportable under 10 CFR 50.73(a)(2)(i)(B)
Cause of Event
The cause of the three SVs being outside of their allowable as-found setpoint is due to setpoint drift. The ASME Code acknowledges setpoint drift by requiring the as-left setpoint to be +/-1% and allowing the as-found setpoint to be +/-3%.
Corrective Actions
The nine SVs removed during the 1 R20 refueling Outage in November 2004 were replaced with refurbished SVs that met the Technical Specification 4.3E requirement of an as-left setpoint tolerance of
+/-1%.
Additional Information
A. Failed Components:
Three Main Steam Line Safety Valves (SVs) were determined to have setpoints out of tolerance.
NRC FORM 366 (U*204)
(If more space is required, use additional copies of NRC Form 366A)
B. Previous Similar Events
Based on further review of this event, it has been determined that during previous operating cycles, Oyster Creek has found similar out of tolerance results from previous laboratory testing. At the time of those discoveries it was determined that those events were not reportable in accordance with the guidance found in NUREG 1022 regarding criteria for not reporting routine surveillance results having no safety consequences. In previous events and this current event, it was determined that the capability of performing the safety functions of the SVs with the out of tolerance setpoints was maintained.
C. Identification of Components referred to in this Report:
Components IEEE 805 System ID IEEE 803A Function Safety Valves Reactor Pressure Vessel E11S-SB EIIS-RCT EIIC-RV EIIC-RPV NRC FORM 366 (6.2004)