05000410/LER-2011-001, Regarding As-Found Safety Relief Valve Lift Setpoints Exceed Technical Specification Allowable Values
| ML11158A064 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 05/31/2011 |
| From: | Lynch T Constellation Energy Group, EDF Development, Nine Mile Point |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 11-001-00 | |
| Download: ML11158A064 (8) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 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 |
| 4102011001R00 - NRC Website | |
text
,jt Thomas A. Lynch P.O. Box 63 Plant General Manager Lycoming, New York 13093 315.349.5205 315.349.1321 Fax CENG.
a joint venture of AConstellation Energ i
T',yeDF NINE MILE POINT NUCLEAR STATION May 31, 2011 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTENTION:
Document Control Desk
SUBJECT:
Nine Mile Point Nuclear Station Unit No. 2; Docket No. 50-410 Licensee Event Report 2011-001, As-Found Safety Relief Valve Lift Setpoints Exceed Technical Specification Allowable Values In accordance with 10 CFR 50.73(a)(2)(i)(B), please find attached Licensee Event Report (LER) 2011-001, As-Found Safety Relief Valve Lift Setpoints Exceed Technical Specification Allowable Values.
Nine Mile Point Nuclear Station, LLC (NMPNS) is currently completing the investigation and evaluation of this event. Upon completion of these actions, NMPNS will submit a supplement to the LER.
Should you have questions regarding the information in this submittal, please contact John J. Dosa, Director Licensing, at (315) 349-5219.
Very truly yours, TAL/DEV
Attachment:
Licensee Event Report 2011-001, As-Found Safety Relief Valve Lift Setpoints Exceed Technical Specification Allowable Values cc:
Regional Administrator, NRC Project Manager, NRC Resident Inspector, NRC
ATTACHMENT LICENSEE EVENT REPORT 2011-001 AS-FOUND SAFETY RELIEF VALVE LIFT SETPOINTS EXCEED TECHNICAL SPECIFICATION ALLOWABLE VALUES Nine Mile Point Nuclear Station, LLC May 31, 2011
i NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 (10-2010)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE Nine Mile Point Unit 2 05000410 1 OF 6
- 4. TITLE As-Found Safety Relief Valve Lift Setpoints Exceed Technical Specification Allowable Values
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR None NA NUMBER NO.NoeA 04 01 2011 2011 001 00 05 31 2011 FACILTYNone DOCKETNUMBERA I
INoeA
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
El 20.2201(b)
[I 20.2203(a)(3)(i)
[I 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii) 1E 20.2201(d)
E] 20.2203(a)(3)(ii)
[I 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
[I 20.2203(a)(1)
[I 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
- 10. POWER LEVEL [I 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
El 50.73(a)(2)(iii)
[: 50.73(a)(2)(ix)(A)
El 20.2203(a)(2)(ii).
[E 50.36(c)(1)(ii)(A)
El 50.73(a)(2)(iv)(A)
[I 50.73(a)(2)(x)
[I 20.2203(a)(2)(iii)
El 50.36(c)(2)
El 50.73(a)(2)(v)(A)
El 73.71(a)(4) 100 [1 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
El 73.71(a)(5)
El 20.2203(a)(2)(v)
[1 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
El OTHER El 20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specify in Abstract below or in NMP2 TS 3.4.4 requires the safety function of sixteen (16) SRVs to be operable in reactor operating modes 1, 2, and 3. With one or more required SRVs inoperable, the unit is required to be placed in Mode 3 (hot shutdown) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 (cold shutdown) within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Since the as-found testing determined that 4 of the 18 SRVs were inoperable for an indefinite period of time during the operating cycle that preceded the 2010 refueling outage, it is probable that NMP2 operated longer than the TS allowed Completion Time.
C. INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:
None
D. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES
April 2010 During the 2010 refueling outage, all 18 SRVs are removed and replaced with pre-tested SRVs that had completed set pressure certification lifts within plus 1 percent to minus 0.5 percent of the specified set pressure.
5/2/2010 NMP2 startup from the 2010 refueling outage commences with replacement SRVs installed.
4/1/2011 NMPNS documents in Condition Report 2011-003045 that the as-found lift pressure for 4 SRVs removed during the 2010 refueling outage exceeded the TS-specified setpoint plus or minus 3 percent.
E. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED
None
F. METHOD OF DISCOVERY
The out-of-tolerance SRV lift setpoints were discovered during the performance of as-found testing conducted at NWS Technologies in Spartanburg, South Carolina.
G. MAJOR OPERATOR ACTION:
None. No operational conditions requiring operator action occurred as a result of this event.
H. SAFETY SYSTEM RESPONSES:
None. No operational conditions requiring the response of safety systems occurred as a result of this event.
NRC FORM 366 (10-2010)
II. CAUSE OF EVENT
The immediate cause for this reportable condition is out-of-tolerance lift pressures that exceeded the TS-allowed values for 4 of 18 SRVs, and which existed for longer than the TS allowed Completion Time. NMPNS is currently completing the investigation and evaluation of this event to determine the cause and any contributing factors.
Ill. ANALYSIS OF THE EVENT:
This event is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as an operation or condition prohibited by the NMP2 TS. NMP2 TS 3.4.4 requires the safety function of 16 SRVs to be operable in reactor operating modes 1, 2, and 3. With one or more required SRVs inoperable, the unit is required to be placed in Mode 3 (hot shutdown) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 (cold shutdown) within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The as-found testing determined the lift pressures for 4 of the 18 SRVs to be outside of the TS requirements. Consistent with the guidance provided in NUREG-1 022, Revision 2, Section 3.2.2 (Example (3), Multiple Test Failures), the condition is considered to have existed during the plant operating cycle preceding the 2010 refueling outage (Cycle 12) and is reportable under 10 CFR 50.73(a)(2)(i)(B).
The ASME Boiler and Pressure Vessel Code requires that the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system, the size and number of SRVs are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary.
The NMP2 SRVs are Dikkers Model G471-6 valves. There are a total of 18 installed SRVs divided into 5 groups, with each group having a different lift pressure setpoint, as follows:
Number of SRVs Setpoint (psiq) 2 1165 +/-35.0 4
1175 +/-35.0 4
1185 +/-36.0 4
1195 +/-36.0 4
1205 +/-36.0 The SRVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. Each SRV discharges steam through a discharge line to a point below the minimum water level in the suppression pool.
The overpressure protection system must accommodate the most severe pressure transient. For NMP2, the most severe transient is the closure of all main steam isolation valves (MSIVs) followed by a reactor scram on high neutron flux (assumes failure of the direct scram associated with MSIV position). The analysis results demonstrate that the design SRV capacity is capable of maintaining reactor pressure below the ASME Code limit of 1375 psig (110 percent of the 1250 psig vessel design pressure). For the purpose of the overpressure protection analysis, 16 of the SRVs with the highest setpoints are assumed to operate in the safety mode (i.e., operation of the two SRVs with NRC FORM 366 (10-2010)
setpoints of 1165 psig is not credited in the analysis), with assumed setpoints that are about 3 percent above the nominal setpoints. Since the SRV with the largest deviation from the TS-required lift pressure has a TS-required setpoint of 1165 psig, the as-found setpoint deficiency for this SRV has no impact on the overpressure protection analysis. The setpoints for the other 3 SRVs with out-of-tolerance lift pressures exceeded the TS required as-found values (including 3 percent tolerance) by only small amounts, which would have a minimal impact on the overpressure protection analysis results. In addition, parametric analyses presented in Appendix 15C of the NMP2 Updated Safety Analysis Report (USAR) demonstrate that if only 14 SRVs are assumed to operate rather than 16 SRVs, the peak calculated vessel pressure increases by less than 10 psig.
Based on the above discussion, the margin between the calculated peak vessel pressure evaluated for Cycle 12 (1301 psig) and the ASME Code limit of 1375 psig, and the fact that all 18 SRVs actually lifted during the as-found testing, the peak vessel pressure would not have exceeded 1375 psig had an overpressure transient occurred that required SRV operation. Furthermore, the peak reactor steam dome pressure would also have remained below the TS safety limit of 1325 psig.
Overpressure analyses for the limiting NMP2 Anticipated Transients Without Scram (ATWS) event (MSIV closure) have also previously been performed to demonstrate that the reactor pressure does not exceed the ASME Service Level C design limit of 1500 psig. For this analysis, two SRVs were assumed to be unavailable. The analysis results, presented in NMP2 USAR Appendix 15C, showed a peak calculated vessel bottom head pressure of 1279 psig. Based on the margin between this calculated value and the 1500 psig limit, the small amount by which the 4 SRVs exceeded the TS setpoint limits, and the fact that all 18 SRVs actually lifted during the as-found testing, the peak vessel pressure would not have exceeded 1500 psig had an ATWS event occurred that required SRV operation.
One of the 4 SRVs with an out-of-tolerance lift pressure (ID No. 2MSS*PSV137; Serial No. 160970) is also an Automatic Depressurization System (ADS) valve. The lift pressure deficiency had no impact on the ADS function of this SRV.
Based on the above, it is concluded that the safety significance of this event is low and the event did not pose a threat to the health and safety of the public or plant personnel.
IV. CORRECTIVE ACTIONS
A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:
During the 2010 NMP2 refueling outage, all 18 of the SRVs were removed and replaced with pre-tested SRVs that had completed set pressure certification lifts within plus 1 percent to minus 0.5 percent of the specified set pressure, thereby meeting the NMP2 TS 3.4.4 requirements.
The affected SRVs will be refurbished, tested, and certified prior to future use at NMP2.
B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:
NMPNS is currently completing the investigation and evaluation and developing associated corrective actions for this event. Identified corrective actions will be provided in a supplement to this LER.
NRC FORM 366 (10-2010)
V. ADDITIONAL INFORMATION
A. FAILED COMPONENTS:
The identified condition for the 4 SRVs is considered to be a Maintenance Rule functional failure since the 4 SRVs did not lift within the setpoint tolerance requirements of the ASME Operation and Maintenance (OM) Code - 2004 Edition.
B. PREVIOUS LERs ON SIMILAR EVENTS:
None C. THE ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) COMPONENT FUNCTION IDENTIFIER AND SYSTEM NAME OF EACH COMPONENT OR SYSTEM REFERRED TO IN THIS LER:
COMPONENT Reactor Pressure Vessel Main Steam Lines Safety Relief Valve D. SPECIAL COMMENTS:
None IEEE 803 FUNCTION IDENTIFIER RPV IEEE 805 SYSTEM IDENTIFICATION AD SB SB RV NRC FORM 366 (10-2010)