05000286/LER-2015-006, Regarding Technical Specification Prohibited Condition Due to Two Pressurizer Code Safety Valves Discovered Outside Their As-Found Lift Setpoint Test Acceptance Criteria
| ML15251A236 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 08/28/2015 |
| From: | Coyle L Entergy Nuclear Operations, Indian Point |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-15-108 LER 15-006-00 | |
| Download: ML15251A236 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(1), Submit an LER, Invalid Actuation 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2862015006R00 - NRC Website | |
text
En tergy Iawienc Poinnryl Cete 450e Bieroaidwaynt Lawrence Coyle Site Vice President NL-1 5-1 08 August 28, 2015 U. S. Nuclear Regulatory Commission Document Control Desk 11545 Rockville Pike, TWFN-2 F1 Rockville, MD 20852-2738
SUBJECT:
Licensee Event Report # 2015-006-00, "Technical Specification Prohibited Condition Due to Two Pressurizer Code Safety Valves Discovered Outside Their As-Found Lift Setpoint Test Acceptance Criteria" Indian Point Unit No. 3 Docket No. 50-286 DPR-64
Dear Sir or Madam:
Pursuant to 10 CFR 50.73(a)(1), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2015-006-00. The attached LER identifies an event where there was a Technical Specification prohibited condition for two inoperable Pressurizer Safety Valves, which is reportable under 10 CFR 50.73(a)(2)(i)(B). This condition was recorded in the Entergy Corrective Action Program as Condition Report CR-I P3-201 5-03710.
There are no new commitments identified in this letter. Should you have any questions regarding this submittal, please contact Mr. Robert Walpole, Manager, Regulatory Assurance at (914) 254-6710.
Sincerely, Lq cc:
Mr. Daniel H. Dorman, Regional Administrator, NRC Region I NRC Resident Inspector's Office, Indian Point Energy Center Ms. Bridget Frymire, New York State Public Service Commission I~J~JL
Abstract
On July 1, 2015, Engineering was notified by Wyle Laboratories that two of three Pressurizer Code Safety Valves (RC-PCV-464 and RC-PCV-468) were outside their As-Found lift set point test acceptance criteria (2411 2559 psig).
The As-Found set pressure testing acceptance criterion for~operability is 2485 +/-3%.
The SVs were removed during the last refueling outage CR0) in the spring of 2015 and sent offsite for testing.
Testing was performed within one year of removal as required by the Inservice Testing Program.
SV RC-PCV-464 lifted at 2573 psig and SV RC-PCV-468 lifted at 2379 psig which is outside their set pressure range.
The remaining SV tested satisfactorily.
All three SVs were found with zero seat leakage.
During the RO all three SVs were removed and replaced with certified pre-tested spare SVs.
The SVs installed during the RO were As-Left tested to 2485 +/-1% with zero seat leakage in accordance with procedure 3-PT-R5A.
Technical Specification (TS) 3.4.10 (Pressurizer Safety Valves),
requires three pressurizer safety valves to be operable with lift settings set at greater than 2460 psig and less than 2510 psig.
TS Surveillance Requirement (SR) 3.4.10.1 requires each PSV to be verified operable in accordance with the Inservice Testing Program.
The most probable cause of SV RC-PCV-464 lifting greater than 3% of its nominal setpoint was setpoint drift.
The most probable cause of RC-PCV-468 lifting below 3% of its nominal setpoint was spring relaxation.
Corrective action will be to perform a valve disassembly and inspection to determine the cause of the failure.
The event had no effect on public health and safety.
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Safety Significance
This event had no effect on the health and safety of the public.
There were no actual safety consequences for the event because there no events that required the pressurizer SRVs.
An evaluation was performed on the potential impact of the condition on accident analysis and best estimate plant response.
The best estimate plant response to applicable accidents and transients would not be impacted by the as-found valve lift conditions.
The non-LOCA analyses that could be impacted by the condition are:
- 1) Loss of load/Turbine Trip (LOL/TT),
- 2)
Loss of normal feedwater/loss of non-emergency AC power (LONF/LOAC),
- 3)
Loss of flow/locked rotor (LOF/LR),
- 4)
Rod withdrawal at power (RWAP),
and 5) Anticipated Transient Without Scram (ATWS)
The PSRVs provide overpressure protection for the Reactor Coolant System (RCS) in conjunction with the Reactor Protection System.
Events such as Loss of Coolant Accidents (LOCA),
Main Steam Line Break (MSLB),
Steam Generator Tube Rupture (SGTR),
and dropped rod evaluated in UFSAR Chapter 14, produce a decrease in RCS pressure and therefore the reactor coolant pressure boundary (RCPB) protection provided by the PSRVs need not be credited.
An assessment of the existing analyses of the applicable non-LOCA transients concluded that the impact of the condition on these non-LOCA transients is either negligible or insignificant.
The UFSAR Chapter 14 analyses remains valid with the as-found PSRV condition.
In the UFSAR Chapter 14 analysis, the opening setpoint of the three PSRVs is assumed to be at +/-4% of the n~ominal 2485 psig value for all four of the Chapter 14 transients (LOL/TT, LONF/LOAC, LOF/LR, and RWAP)
PSRV RC-PCV-464 as-found lift pressure was 2573 (3.54% above nominal setpoint of 2485 psig).
PSRV RC-PCV-468 as-found lift pressure was 2379 psig (4.26% below nominal setpoint pressure of 2485 psig).
The as-found result of the PSRV RC-PCV-464 test
(+3.54%)
is acceptable from a safety function standpoint since
+3.54% is less than +4%.
The as-found result of the PSRV RC-PCV-468 test
(-4.26%)
is acceptable from a pressure boundary protection safety function standpoint since 4.26% is well below +4%.
For the LONF/LOAC event the concern is not only RCS over pressurization but potential water relief from a solid pressurizer.
The analysis of the LONF/LOAC event conservatively assumes that the pressurizer pressure control system (including pressurizer spray and PORVs) is available and that the PSRV setpoint is
- - 4% of nominal.
This assumption involving the PSRVs is conservative because it would allow earlier mass release from the pressurizer if water solid conditions should occur but has the potential undesirable result of loss of significant RCS mass and volume.
The analysis of the LONF/LOAC transient for the Stretch Power Uprate (SUP)
Project showed the pressurizer would not become water solid because of acceptable function of the Auxiliary Feedwater system and the Main Steam Safety Valves and/or Main Steam Atmospheric Dump Valves.
Therefore, the early lift of PSRV RC-PCV-468 at a setting slightly below -4% of nominal has no bearing on the progression of the LONE/LOAC event.
Although the anticipated transient without scram (ATWS) is not one of the transients evaluated in UFSAR Chapter 14, the PSRVs are credited for this event.
The design capacity of each of the three PSRVs is 420,000 ibm/hr.
This capacity is greater than the PSRV relief capacity of 408,000 Ibm/hr assumed in the 1979 generic ATWS analysis.
As such, this would result in an overall peak pressure benefit when compared to peak RCS pressure calculated for the generic limiting ATWS events.
Further, the most current peak RCS pressure determined in the most recent ATWS analysis for the Unit 3 Stretch Power Uprate Project is 2862 psia.
This peak RCS pressure, the TS safety limit (SL)
, and the as-found lift pressures are all significantly less than the ASME B&PV Code Level C service stress criterion input of 3215 psia.