05000277/LER-2010-003, Regarding Laboratory Analysis Identifies Safety Relief Valves and Safety Valve Set Point Deficiencies
| ML103260460 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  (DPR-044) |
| Issue date: | 11/19/2010 |
| From: | Stathes G Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 10-003-00 | |
| Download: ML103260460 (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 |
| 2772010003R00 - NRC Website | |
text
Exelon.
Exelon Nuclear www.exeloncorp.com Nuclear Peach Bottom Atomic Power Station 1848 Lay Rd.
Delta, PA 17314 10CFR 50.73 November 19, 2010 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Peach Bottom Atomic Power Station (PBAPS) Unit 2 Facility Operating License No. DPR-44 NRC Docket No. 50-277
Subject:
Licensee Event Report (LER) 2-10-03 Enclosed is a Licensee Event Report concerning a condition prohibited by Technical Specifications involving two Safety Relief Valves (SRVs) and one Safety Valve (SV) that did not meet their Technical Specification + 1% set point tolerance when tested in the laboratory.
In accordance with NEI 99-04, the regulatory commitment contained in this correspondence is to restore compliance with the regulations. The specific methods that are planned to restore and maintain compliance are discussed in the LER. If you have any questions or require additional information, please do not hesitate to contact us.
Sincerely, Garey L. Stathes Plant Manager Peach Bottom Atomic Power Station GLS/djf/cee/IR 1120516 Attachment cc:
US NRC, Administrator, Region I US NRC, Senior Resident Inspector R. R. Janati, Commonwealth of Pennsylvania S. Grey, State of Maryland P. Steinhauer, PSE&G, Financial Controls and Co-owner Affairs INPO Records Center CCN: 10-94
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 (10-2010)
, the NRC may sfor each block) not conduct or sponsor, and a person is not required to respond to, the digits/characters finformation collection.
- 13. PAGE Peach Bottom Atomic Power Station (PBAPS) Unit 2 05000277 1 OF 4
- 4. TITLE Laboratory Analysis Identifies Safety Relief Valves and Safety Valve Set Point Deficiencies
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED jFACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEARI 05000 NUMBER NO.
IFACILITY NAME DOCKET NUMBER 09 27 2010 10 003-00 11 19 2010 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
El 20.2201(b)
El 20.2203(a)(3)(i)
[I 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii)
El 20.2201(d)
El 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
[I 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
[I 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
[] 20.2203(a)(2)(i)
[I 50,36(c)(1)(i)(A)
El 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL El 20.2203(a)(2)(ii)
[I 50.36(c)(1)(ii)(A)
El 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
El 20.2203(a)(2)(iii)
[I 50.36(c)(2)
[I 50.73(a)(2)(v)(A)
El 73.71(a)(4) 0%
El 20.2203(a)(2)(iv)
E] 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
El 73.71(a)(5)
El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
[E 50.73(a)(2)(v)(C)
El OTHER El 20.2203(a)(2)(vi)
ED 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specify in Abstract below or in
Analysis of the Event
There were no actual safety consequences associated with this event.
This report is being submitted pursuant to:
10CFR 50.73(a)(2)(i)(B) - Condition Prohibited by Technical Specifications - Technical Specification Limiting Condition for Operation (LCO) 3.4.3 requires that 11 of the 13 SRVs / SVs be operable during operational Modes 1, 2, and 3.
Contrary to this requirement, two SRVs and one SV were found with set points outside of the Technical Specification requirements.
The ASME Boiler and Pressure Vessel Code requires that Reactor Pressure Vessel (EIIS:
RCT) be protected from overpressure during upset conditions by self-actuated relief valves. As part of the nuclear pressure relief system, the size and number of SRVs and SVs are selected such that the peak pressure in the nuclear system will not exceed the ASME Code limits for the Reactor Coolant Pressure Boundary. There exists a total of 13 SRVs / SVs installed on the four Main Steam (EIIS: SB) Lines. The eleven installed SRVs exhaust steam through discharge lines to a point below the minimum water level in the Suppression Pool. The two installed SVs discharge steam directly to the Drywell. The SRVs and SVs are located on the four main steam lines (EIIS: SB) within Primary Containment. The SRVs are 'three-stage' valves consisting of a main valve disc and piston (third stage) operated by a second stage disc and piston displaced by either a first stage pressure-sensing pilot (for overpressure protection) or a pneumatically-operated mechanical push rod (for the ADS function or for remote-manual operation). The SVs are direct-acting, spring loaded relief valves.
During the Unit 2 Cycle 18 operations, there were no plant transients that required automatic or manual SRV / SV operation. The as-found set points for the three SRVs / SV that tested outside of their Technical Specification allowable range were high. There were a total of six SRVs and one SV removed for testing and replacement during the 1 8th Refueling Outage. All three SRVs / SV outside of their Technical Specification allowable range were within the ASME Code allowable +/- 3% tolerance. An engineering analysis determined one of the three SRVs /
SV found outside of its Technical Specification range drifted into the acceptance range of a second SRV that had drifted out of its range. The net plant impact found that only 2 of 13 SRVs
/ SVs lifted outside the bounds of Tech Spec ranges (i.e. +/-1%). Therefore, 11 of 13 remained within the lift pressure range as specified by Tech Specs. One of the two SRVs (SRV S/N 182) was also an Automatic Depressurization System (ADS) valve. The set point drift had no impact on the ADS or manual function of the valves.
The event is not considered to be risk significant.
Corrective Actions
The two SRVs and the one SV were replaced with refurbished SRVs / SV for the 1 9 th Unit 2 operating cycle.
A change to the PBAPS licensing basis (e.g., extended power uprate) will be pursued to increase SRV / SV set point tolerances.
Previous Similar Occurrences There were three previous LERs identified involving SRVs / SVs exceeding their Technical Specification +/- 1% set point requirement. LER 3-07-01 reported three SRVs / SV (two SRVs and one SV) having its as-found set points in excess of the Technical Specification allowable +/-
1% tolerance.
LER 2-06-02 reported one SV having its as-found set points in excess of the Technical Specification allowable +/- 1% tolerance. LER 3-05-04 reported a situation involving four SRVs having their as-found set points in excess of the Technical Specification allowable +/-
1% tolerance. The previous SRV / SV as-found set points were all within the +/- 3% ASME code allowable set point tolerance. Completed corrective actions addressing set point drift for these previous events involved replacement of the previous SRVs with different SRVs and therefore, would not have been expected to prevent this event.
One of the SRVs / SV reported in this LER (2-10-003) was found in the same location as those previously reported SRVs / SV in LER 2-06-02. However, the serial number of the valve is not the same. During P2R16, the 70A SV was determined to have its as-found set points in excess of the Technical Specification allowable +/- 1% tolerance (but well within the ASME Code allowable +/- 3% tolerance).
The as-found set point was 1.2% higher than the Technical Specification allowable. During P2R18, the 70A SV was determined to have an as-found set point of 1.81% higher than the Technical Specification allowable. Neither of the other two SRVs reported in this LER (2-10-003) were the same as these previously reported in LERs 2-06-02, 3-05-04, and 3-07-01.