05000315/LER-2019-001, Two Main Steam Safety Valve Setpoints Found Outside Technical Specification Limits
| ML19122A171 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 04/30/2019 |
| From: | Lies Q Indiana Michigan Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| AEP-NRC-2019-15 LER 2019-001-00 | |
| Download: ML19122A171 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) |
| 3152019001R00 - NRC Website | |
text
m INDIANA MICHIGAN POWER A unit of American Electric Power April 30, 2019 Docket No.: 50-315 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike Rockville, MD 20852 Donald C. Cook Nuclear Plant Unit 1 LICENSEE EVENT REPORT 315/2019-001-00 Indiana Michigan Power Cook Nuclear Plant One Cook Place Bridgman, Ml 49106 Indiana Michigan Power.com AEP-NRC-2019-15 10 CFR 50.73 Two Main Steam Safety Valve Setpoints found outside Technical Specifications Limits In accordance with 10 CFR 50.73, Licensee Event Report (LER) System, Indiana Michigan Power Company, the licensee for Donald C. Cook Nuclear Plant Unit 1, is submitting as an enclosure to this letter the following report:
LER 315/2019-001-00: Two Main Steam Safety Valve Setpoints found outside Technical Specifications Limits There are no commitments contained in this submittal.
Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Director, at (269) 466-2649:
Sincerely, l~j~
Site Vice President MPH/mll
Enclosure:
Licensee Event Report 315/2019-001-00: Two Main Steam Safety Valve Setpoints found outside Technical Specifications Limits
U.S. Nuclear Regulatory Commission Page 2 c:
R. J. Ancona - MPSC R. F. Kuntz - NRG Washington DC MDEQ - RMD/RPS NRG Resident Inspector D. J. Roberts - NRG Region Ill A. J. Williamson - AEP Ft. Wayne AEP-NRC-2019-15
Enclosure to AEP-NRC-2019-15 Licensee Event Report 315/2019-001-00 Two Main Steam Safety Valve Setpoints ~ound outside Technical Specifications Limits
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2018)
Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.
i~~olA*~.
LICENSEE EVENT REPORT (LER)
Reported lessons learned are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S.
-....;~.. /#
. (See Page 2 for required number of digits/characters for each block)
Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory (See NUREG: 1022, R.3 for instruction arid guidance for completing this form Affairs, NEOB-10202, (3150--0104), Office of Management and Budget, Washington, DC 20503. ~
a means used to impose an information collection does not display a currently valid 0MB control htt12://www.nrc.gov/reading-rm/doc-collections/nu regs/staff/sr1 022lr3D number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
,3. PAGE Donald C. Cook Nuclear Plant Unit 1 05000315 1 OF4
- 4. TITLE Two Main Steam Safety Valve Setpoints Found Outside Technical Specification Limits
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED YEAR ISEQUENTIAi REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NO.
MONTH DAY YEAR NIA 05000 NUMBER FACILITY NAME DOCKET NUMBER 03 04 2019 2019 001 00 04 30 2019 NIA 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 1 D 20.2201(b)
D 20.2203(a)(3)(i)
D 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A)
D 20.2201 (d)
D 20.2203(a)(3)(ii)
D 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B)
D 20.2203(a)(1)
D 20.2203(a)(4)
D so.73(a)(2)(iii)
D 50.73(a)(2)(ix)(A)
D 20.2203(a)(2)(i)
D 50.36(c)(1)(i)(A)
D 50.73(a)(2)(iv)(A)
D 50.73(a)(2)(x)
- 10. POWER LEVEL D 20.2203(a)(2)(ii)
D 50.36(c)(1 )(ii)(A)
D 50.73(a)(2)(v)(A)
D 73.71(a)(4)
D 20.2203(a)(2)(iii)
D so.36(c)(2)
D 50.73(a)(2)(v)(B)
D 73.71(a)(5) 49 D 20.2203(a)(2)(iv)
D so.46(a)(3)(ii)
D 50.73(a)(2)(v)(C)
D 73.77(a)(1)
D 20.2203(a)(2)(v)
D 50.73(a)(2)(i)(A)
D 50.73(a)(2)(v)(D)
D 73.77(a)(2)(i)
D 20.2203(a)(2)(vi)
~ 50.73(a)(2)(i)(B)
D 50.73(a)(2)(vii)
D 73.77(a)(2)(ii)
D 50.73(a)(2)(i)(C) 00THER Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER LICENSEE CONTACT rELEPHONE NUMBER (Include Area Code)
Michael K. Scarpello, Regulatory Affairs Director (269) 466-2649 '
MANUS REPORTABLE MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT FACTURER TO ICES
CAUSE
SYSTEM COMPONENT FACTURER TOEPIX X
SB RV Dresser y
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR 0 YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
IXJ NO SUBMISSION DATE
~BSTRACT (Limit lo 1400 spaces, ie.* approximately 14 single-spaced typewritten lines)
On March 4, 2019, Donald C. Cook Unit 1 was in Mode 1 and holding at approximately 49% reactor power while performing Technical Specification (TS) surveillance testing on the Main Steam Safety Valves (MSSVs) setpoints, just prior to the Unit 1 Cycle 29 (U1 C29) Refueling Outage.
The Unit 1 Steam Generator (SG) #2 MSSVs 1-SV-1 B-2 and 1-SV-2A-2 as found setpoints were above the TS allowable setpoint pressure of +/- 3% lift setpoint pressure.
Based on the guidance in NUREG-1022, Revision 3, the Unit 1 SG #2 MSSVs were considered to be inoperable longer than allowed by TS.
The cause of the failure was determined to be setpoint drift. The MSSV setpoints were adjusted and, after adjustment, the MSSVs passed the as left testing criteria of +/- 1 % lift setpoint pressure. Since a similar cause was determined for both unsatisfactory as found lift pressures, this condition may have arisen over a period of time, and there is a likelihood that the affected MSSVs may not have been operable during plant operation.
Therefore, this event is reportable as an Operation or Condition Prohibited by TS in accordance with 10 CFR 50. 73(a)(2)(i)(B).
NRC FORM 366 (04-2018)
EVENT DESCRIPTION
SEQUENTIAL NUMBER
- - 001 REV NO.
- - 00 On March 4, 2019, Donald C. Cook Unit 1 was in Mode 1 and holding at approximately 49% reactor power while performing Technical Specification (TS) surveillance testing on the Main Steam Safet_y V~lves (MSSVs) [SB] °[RVf setpoints, just prior fo the Unit f Cycie 29 (U1C29) Refue-ling Outage.
The as found setpoint pressure of the Unit 1 Steam Generator (SG) [SG] #2 MSSV 1-SV-18-2 was 1109.96 pounds per square inch gauge (psig), which was 4.22% above the TS allowable setpoint pressure of +/- 3% of 1065 pj:iig. The valve passed the seat-leakage requirement. After adjustment, the valve passed the as left testing criteria of +/- 1 % of the setpoint of 1065 psig.
The as found setpoint pressure of the Unit 1 SG #2 MSSV 1-SV-2A-2 was 1117.32 psig, which was 3.94%
above the ts ailowable setpoint pressure of +/~ 3% of 1075 psig. The valve passed the seat-leakage requirement.
After adjustment, the valve passed the as left testing criteria of +/- 1 % of the setpoint of 1075 psig.
As a result of the failure of 1-SV-1 B-2 and 1-SV-2A-2, two additional valves for each failure (four valves in all) were tested as required by the ASME OM Code.
All four of the additional valves passed the as found testing within the required criteria of +/- 3% for the setpoint test. The three other MSSVs protecting the Unit 1 SG #2 were also tested as part of the original scope, and all passed the as found test within the required setpoint criteria of+/- 3%.
T$ Limiting Co11diJion For Operation 3,7.1 requires a min_imurn of 20_M_SSVs (5 per $G) t_o be_ operable in Modes 1, 2, and 3. Since a similar cause was determined for both unsatisfactory as found lift pressures, this condition may have arisen over a period of time, and there is a likelihood that the affected MSSVs may not have been operable during plant operation. Therefore, this occurrence is considered reportable as an Operation or Condition Prohibited by TS in accordance with 10 CFR 50. 73(a)(2)(i)(B).
COMPONENT 1-SV-18-2 & 1-SV-2A-2 are Consolidated-Dresser model 3707RA Valves Page 2 of 4 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2018)
Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.
LICENSEE EVENT REP
- R)
Reported lessons learned are incorporated into the licensing process and fed back to industry.
- ORT (LE Send comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear CONTINUATION SHEET Regulatory.
Commission, Washington, DC 20555-0001, or by e-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of lnfonmation and Regulatory 1-------------------------1 Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a
{See NUREG-1022, R.3 for. instruction and guidance for completing this form http://www'.nrc.gov/reading~rm/doc-collections/nuregs/staff/sr1022/r3D means used to impose an infonmation* collection does not display a currently valid OMS.control number, the NRG may not conduct or sponsor, and a person is not required to respond to, the infonmation.collection.
- 3. LER NUMBER Donald C. Cook Nuclear Plant Unit 1 05000315
CAUSE OF THE EVENT
APPARENT CAUSE
.YEAR 2019 SEQUENTIAL NUMBER
- - 001 REV NO.
- - Ob The apparent cause of the failed as found set-pressure testing on both 1-SV-18-2 and 1-SV-2A-2 was setpoint drift. Historical analysis of 122 previous as found tests showed that 72 experienced a rise in the as found setpoint *while 50 *valves experienced a decrease* iri as found setp6int. Of the 122 previous as found tests, there have been five failures. In addition, over the course of 10 years, many valves have experienced both an increase and a decrease in as found setpoint from one test to the next.
This indicates that choosing a setpoint above or below the nominal value would not necessarily decrease the likelihood of failure. Therefore, no further corrective action is necessary because of the low frequency of failure of these valves and the historical randomness of setpoint drift.
ASSESSMENT OF SAFETY CONSEQUENCES
NUCLEAR SAFETY There was no actual or potential nuclear safety hazard resulting from the Two Main Steam Safety Valve Setpoints being Found Outside Technical Specification Limits
- INDUSTRIAL SAFETY There was no actual or potential industrial safety hazard resulting from the Two Main Steam Safety Valve Setpoints being Found Outside Technical Specification Limits
. RADIOLOGICAL SAFETY.
There was no actual or potential radiological safety hazard resulting from the Two Main Steam Safety Valve Setpoints being Found Outside Technical Specification Limits PROBABILISTIC RISK ASSESSMENT (PRA) 1-SV-18-2 failed because it lifted high at 1109.96 psig, which is above its setpoint of 1065 psig. A similar valve on the sa*me steam generator, 1-SV-2A-2 also failed high at 1117.32 psig, which is above its setpoint of 1075 psig. The only explicitly modeled function of these valves in the PRA models is to remain closed following core damage to prevent the release of contaminants following a tube rupture event. These valves failing to lift at their appropriate setpoint does not impact the ability of 1-SV-2A-2 and 1-SV-18-2 to meet their PRA-modeled function.
For this reason, this condition does not impact any of the PRA estimations for Core Damage Frequency (GDF) or Large Early Release Frequency (LERF).
Qualitatively, these valves are responsible for protecting Unit 1 SG #2 from damage.due to an overpressure event. Each SG has five of these MSSVs and one Power-Operated Relief Valve (PORV).
- NRG FORM 366A (04-2018) :
. Page 3 of 4 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2018)
. LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.
Reported -lessons learned are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of lnfonmation and Regulatory 1---------------------------1 Affairs, NEOB-10202, (3150.-0104), Office of Management and Budget, Washington, DC 20503. If a (See NUREG-1022, R.3 for instruction and guidance for completing this form http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3D means used to impose an information collection does not display a currently valid 0MB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. LERNUMBER YEAR Donald C. Cook Nuclear Plant Unit 1 05000315 2019
- SEQUENTIAL NUMBER
- - 001 REV NO.
- - 00 Because these valves serve only to mitigate an overpressure event following an increase in SG pressure to the MSSV setpoint, their failure to lift does not contribute to the frequency of occurrence of these overpressure events. If an overpressure event were to occur, it could be mitigated by a combination of the remaining three MSSV's on Unit 1 SG #2 or by operators using the SG PORV to reduce pressure. The success of any of these mitigation systems would result in a plant trip, but no further complications to plant response. Due to the redundancy present in overpressure mitigation components, it is unlikely that an
Therefore, it can be concluded that this event was of very low safety significance.
CORRECTIVE ACTIONS
Safety valves 1-SV-1 B-2 and 1-SV-2A-2 were successfully adjusted and subsequently,passed the as left testing criteria of +/- 1 % of the setpoint Safety valves 1-SV-1A-3, 1-SV-2A-3, 1-SV-3-3, and 1-SV-1A-1 were tested as expanded scope valves.
All four valves passed their as found tests.
PREVIOUS SIMILAR EVENTS
LERs for Donald G. Cook Nuclear Plant Unit 1 and Unit 2 were reviewed for the previous five years and found no similar events.
- NRG FORM 366A (04-2018)
- Page 4 of 4