05000278/LER-2007-001, Laboratory Analysis Identifies Safety Relief Valves and Safety Valve Set Point Deficiencies
| ML073400422 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 11/26/2007 |
| From: | Massaro M Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 07-001-00 | |
| Download: ML073400422 (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)(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 |
| 2782007001R00 - NRC Website | |
text
Exelkn Exelon Nuclear www.exeloncorp.com Peach Bottom Atomic Power Station Nucear 1848 Lay Road Delta, PA 17314-9032 10CFR 50.73 November 26, 2007 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Peach Bottom Atomic Power Station (PBAPS) Unit 3 Facility Operating License No. DPR-56 NRC Docket No. 50-278
Subject:
Licensee Event Report (LER) 3-07-01 This LER reports 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. There are no regulatory commitments contained in the LER.
If you have any questions or require additional information, please do not hesitate to contact us.
Sincerely, Michael J. Massaro Plant Manager Peach Bottom Atomic Power Station MJM/djf/IR 680967/653733/654019 Attachment cc:
PSE&G, Financial Controls and Co-owner Affairs R. R. Janati, Commonwealth of Pennsylvania INPO Records Center S. Collins, US NRC, Administrator, Region I R. I. McLean, State of Maryland US NRC, Senior Resident Inspector CCN 07-106
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2010 (9-2007)
, the NRC may digits/characters for each block) not conduct or sponsor, and a person is not required to respond to, the infnrmatinn cnllp.rntinn.
- 3. PAGE Peach Bottom Atomic Power Station Unit 3 05000278 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 TSQEI-Lý E
ACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR QU MONTH DAY YEAR C
N05000 NUMBER NO.
6_
oo I oocNU__ooo tFACILITY NAME DOCKET NUMBER 10 4
2007 07
- - 01 00 11 26 20071 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
[l 20.2201(b)
El 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii)
[1 20.2201(d)
El 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
El 20.2203(a)(4) 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B) 0] 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
El 50.73(a).(2)(iii)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL [I 20.2203(a)(2)(ii)
[I 50.36(c)(1)(ii)(A)
[: 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
[I 20.2203(a)(2)(iii)
[I 50.36(c)(2)
[I 50.73(a)(2)(v)(A)
El 73.71(a)(4) 100 [1 20.2203(a)(2)(iv)
[: 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
[1 73.71 (a)(5)
El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
[I 50.73(a)(2)(v)(C)
[I OTHER El 20.2203(a)(2)(vi)
[
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.
The ASME Boiler and Pressure Vessel Code requires that the 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. The 11 installed SRVs exhaust steam through discharge lines to a point below the minimum water level in the Suppression Pool.
The 2 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 remote-manual operation).
The SVs are direct-acting, spring loaded relief valves.
During Unit 3 Cycle 16 operations, there were no plant transients that required automatic SRV
/ SV operation. The as-found set points for the three SRVs / SVs that tested outside of their Technical Specification allowable range were slightly high. There were a total of six SRVs and one SV removed for testing and replacement during the 16th Refueling Outage. All three SRVs
/ SV outside of their Technical Specification allowable range were within the ASME Code allowable of +3%. An analysis determined that the aggregate as-found set point results were bounded by existing analyses, and therefore, there was no significant impact on the PBAPS design basis analyses for the cycle. One of the two SRVs (i.e. SRVs S/N 19) was also an Automatic Depressurization System (ADS) valve. The set point drift had no impact on the ADS or manual function of the valves.
This event is not. considered risk significant.
Cause of the Event
The cause of the SRVs / SV being outside of their allowable as-found set points is due to set point drift. A historical review of SRV as-found test set points indicates that approximately 20% of valves tested over time do not meet the + 1% Technical Specification set point.
Corrective Actions
The two SRVs and the one SV were replaced with refurbished SRVs /SV for the 17 th Unit 3 operating cycle.
To be more consistent with industry practices, changes to the PBAPS licensing basis will be considered to allow for SRV / SV setpoint tolerances of + 3% as allowed by the ASME code.
Previous Similar Occurrences There were two previous LERs identified involving SRVs / SVs exceeding their Technical Specification + 1% setpoint requirement. 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.
LER 2-06-02 reported one SV having its as-found set points in excess of the Technical Specification allowable + 1% tolerance. The previous SRV / SV as-found setpoints were all within the + 3% ASME code allowable setpoint tolerance.
Corrective actions
addressing setpoint drift for these previous events were to replace the SRVs with different
- SRVs, None of the SRVs / SV reported in this LER (3-07-01), were the same as these previously reported SRVs / SV in LERs 3-05-04 and 2-06-02.PRINTED ON RECYCLED PAPERPRINTED ON RECYCLED PAPER