ML16133A170: Difference between revisions

From kanterella
Jump to navigation Jump to search
(Created page by program invented by StriderTol)
(Created page by program invented by StriderTol)
 
Line 1: Line 1:
{{Adams
#REDIRECT [[05000266/LER-2016-001]]
| number = ML16133A170
| issue date = 05/12/2016
| title = LER 16-001-00 for Point Beach, Unit 1, Regarding Degraded Condition
| author name = McCartney E
| author affiliation = NextEra Energy Point Beach, LLC
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000266
| license number = DPR-024
| contact person =
| case reference number = NRC 2016-0019
| document report number = LER 16-001-00
| document type = Letter, Licensee Event Report (LER)
| page count = 3
}}
 
=Text=
{{#Wiki_filter:ENERGY May 12, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Unit 1 Docket 50-266 Renewed License No. DPR-24 Licensee Event Report 266/2016-001-00 Unit 1 Degraded Condition POINT BEACH NRC 2016-0019 10 CFR 50.73 Enclosed is Licensee Event Report (LER) 266/2016-001-00 for Point Beach Nuclear Plant, Unit 1. NextEra Energy Point Beach, LLC is providing this LER regarding the degraded condition on Unit 1. This letter contains no new regulatory commitments.
If you have any questions please contact Mr. Bryan Woyak, Licensing Manager, at 920/755-7599.
Very truly yours, NextEra Energy Point Beach, LLC Eric McCartney Site Vice President Enclosure cc: Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241 NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017 (02-2014)
Estimated burden per response to comply with this mandatory collection request 80 hours.
Reported lessons learned are incorporated into the licensing process and fed back to industry. " Send comments regarding burden estimate to the FOIA, Privacy and lnfonnation Collections Branch & (T*5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by . LICENSEE EVENT REPORT (LER) internet e-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and '?-s-'1 .......... (See Page 2 for required number of Regulatory Affairs, NEOB-1 0202, (3150-01 04), Office of Management and Budget, Washington, DC 20503. If a means used to impose an infonnation collection does not display a currently valid OMB digits/characters for each block) control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
: 1. FACILITY NAME 2. DOCKET IIIIIMRI=R
: 3. PAGE Point Beach Nuclear Plant Unit 1 05000266 1 OF 2 4. TITLE Unit 1 Degraded Condition
: 5. EVENT DATE 6. LER NUMBER 7. REPORT DATE 8. OTHER FACILITIES INVOLVED SEQUENTIAL I Rev FACti tTY NAME UUvl\t:l NUMBER MONTH DAY YEAR YEAR NUMBER NO. MONTH DAY YEAR NA NA FACILITY NAME DOCKET NUMBER 03 15 2016 2016-001 -00 05 12 2016 NA NA 9. OPERATING
: 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) MODE D 20.2201(b)
D 20.2203(a)(3)(i)
D 50.73(a)(2)(i)(C)
D 50.73(a)(2)(vii)
D 20.2201(d)
D 20.2203(a)(3)(ii)
IZl 50.73(a)(2)(ii)(A)
D 50. 73(a)(2)(viii)(A)
MODE5 D D D D 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B)
D 20.2203(a)(2)(i)
D 50.36(c)(1)(i)(A)
D 50. 73(a)(2)(iii)
D 50. 73(a)(2)(ix)(A)
: 10. POWER LEVEL D 20.2203(a)(2)(ii)
D 50.36(c)(1
)(ii)(A) D 50.73(a)(2)(iv)(A)
D 50.73(a)(2)(x)
D 20.2203(a)(2)(iii)
D 50.36(c)(2)
D 50.73(a)(2)(v)(A)
D 73.71 (a)(4) D 20.2203(a)(2)(iv)
D 50.46(a)(3)(ii)
D 50.73(a)(2)(v)(B)
D 73.71 (a)(5) 0% D D D D 20.2203(a)(2)(v) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(C)
OTHER D 20.2203(a)(2)(vi)
D 50.73(a)(2)(i)(B)
D 50.73(a)(2)(v)(D) below or in 12.
CONTACT FOR "HIS LER FACILITY NAME TELEPHONE NUMBER (Include Area Code) Thomas P. Schneider, Senior Licensing Engineer 920-755-7797
: 13. COMPLETE ONE LINE FOR EACH = 'I= LUI DESCRIBED IN THS REPORT CAUSE SYSTEM COMPONENT MANU-REPORTABLE CAUSE SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX FACTURER TO EPIX 8 C8 v C635 y NA NA NA NA NA
: 15. EXPECTED MONTH DAY YEAR SUBMISSSION YES (If yes, complete 15. EXPECTED SUBMISSION DATE) X NO DATE NA NA NA ABSTRACT (Limit to 1400 spaces i.e, approximately 15 single-spaced typewritten lines) At 0649 on March 15, 2016 with Unit 1 shut down in MODE 5 for refueling activities, a boric acid indication upstream of the valve seating surface on the inlet of the valve body of 1 CV-2008, Letdown Orifice 8 Outlet Control Valve was identified as a through-wall flaw. The flaw location was within the reactor coolant system (RCS) pressure boundary as defined by 10 CFR 50.2, "Definitions." The valve body is original plant equipment.
This event is being reported pursuant to 10 CFR 50. 73(a)(2)(ii)(A) for material defects in the primary coolant system that were not acceptable in accordance with ASME Section XI.
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017 (02-2014)
LICENSEE EVENT REPORT (LER) CONTINUATION SHEET 1. FACILITY NAME 2. DOCKET Point Beach Nuclear Plant Unit 1 05000266 NARRATIVE Description of the Event: Estimated burden per response to comply with this mandatory collection request: 80 hours. Reported lessons learned are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate to the FOIA, Privacy and lnfonmation Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555*0001,or byintemet e-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-1 0202, (3150-01 04), Office of Management and Budget, Washington, DC 20503. If a means used to impose an infonmation collection does not display a currently valid OMS control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the infonmation collection.
: 6. LERNUMER YEAR I SEQUENTIAL I NUMBER 2016 -001 REV NO. 00 2 3. PAGE OF 2 At 0649 on March 15, 2016 with Unit 1 shut down in MODE 5 for refueling activities, a boric acid indication upstream of the valve seating surface on the inlet of the valve body of 1 CV-2008, Letdown Orifice B Outlet Control Valve was identified as a through-wall flaw. The flaw location was within the reactor coolant system (RCS) pressure boundary as defined by 10 CFR 50.2, "Definitions." The valve body is original plant equipment.
The Unit 1 Letdown Orifice B Outlet Control Valve (1 CV-2008) has been replaced and returned to service. This 60 day licensee event report is being submitted in accordance with the requirements of 10 CFR 50. 73(a)(2)(ii)(A) for any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being degraded.
The material defect in the primary coolant system could not be found acceptable in accordance with ASME Section XI, IWB-3600, "Analytical Evaluation of Flaws" or ASME Section XI, Table IWB-3410-1, "Acceptance Standards." Cause of the Event: The most likely cause of the degraded barrier was localized corrosion near a sand inclusion indication in the valve body. Analysis of the Event: Maintenance activities identified a through-wall flaw on the valve body of the Letdown Orifice B Outlet Control Valve. The flaw location was upstream of the valve seating surface, on the inlet of the valve body. 1 CV-2008 is one of three valves in parallel in the same application.
The valve was within the RCS pressure boundary, which is defined as being connected to the RCS, up to and including the outermost containment isolation valve in system piping which penetrates the primary reactor containment.
A causal evaluation determined the through-wall flaw in the valve body to most likely be the result of localized corrosion near a sand inclusion indication.
The indication was identified during the manufacturing examination and testing processes.
The indication was accepted by the manufacturer using the acceptance criteria of ASTM Specification E71, "Reference Radiographs for Steel Casting up to 2 inches in Thickness." The localized corrosion most likely allowed system contents to leech through the sand inclusion.
The valve body is original plant equipment and has been in service the entire life of the plant. The flaw has been corrected by replacement of the valve body. Corrective Actions: The Unit 1 Letdown Orifice B Outlet Control Valve (1 CV-2008) has been replaced and returned to service. Destructive testing will be performed on the removed valve to validate the apparent cause. Additional actions/reviews will be taken as required following receipt of the destructive testing report. Examinations have been completed as part of the extent of condition on three valves from the same heat number that had manufacturing indications.
A corrective action has been created to perform a visual examination of one additional valve during the next refueling outage. Safety Significance:
The event was determined to be of very low safety significance.
1 CV-2008 can be isolated from the RCS by a remotely operated valve 1 RC-427, Reactor Coolant Loop B Cold Leg to Charging and Volume Control System (CVCS) Letdown Isolation Valve.
Components with similar sand inclusion indications have been or will be visually examined for evidence of boric acid indications.
While the condition was present, there was no loss of any safety systems, structures or components needed to shut down the reactor, maintain safe shutdown conditions, remove residual heat, control the release of radioactive material or mitigate the consequences of an accident.
There was no impact on the health and safety of the public as a result of this condition.
Similar Events: There have not been similar events of this degraded condition in the past three years. Component Failure Data: 2 INCH-1500 LB COPES-VULCAN MODEL 63 GLOBE VALVE ENERGY May 12, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Unit 1 Docket 50-266 Renewed License No. DPR-24 Licensee Event Report 266/2016-001-00 Unit 1 Degraded Condition POINT BEACH NRC 2016-0019 10 CFR 50.73 Enclosed is Licensee Event Report (LER) 266/2016-001-00 for Point Beach Nuclear Plant, Unit 1. NextEra Energy Point Beach, LLC is providing this LER regarding the degraded condition on Unit 1. This letter contains no new regulatory commitments.
If you have any questions please contact Mr. Bryan Woyak, Licensing Manager, at 920/755-7599.
Very truly yours, NextEra Energy Point Beach, LLC Eric McCartney Site Vice President Enclosure cc: Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241 NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017 (02-2014)
Estimated burden per response to comply with this mandatory collection request 80 hours.
Reported lessons learned are incorporated into the licensing process and fed back to industry. " Send comments regarding burden estimate to the FOIA, Privacy and lnfonnation Collections Branch & (T*5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by . LICENSEE EVENT REPORT (LER) internet e-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and '?-s-'1 .......... (See Page 2 for required number of Regulatory Affairs, NEOB-1 0202, (3150-01 04), Office of Management and Budget, Washington, DC 20503. If a means used to impose an infonnation collection does not display a currently valid OMB digits/characters for each block) control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
: 1. FACILITY NAME 2. DOCKET IIIIIMRI=R
: 3. PAGE Point Beach Nuclear Plant Unit 1 05000266 1 OF 2 4. TITLE Unit 1 Degraded Condition
: 5. EVENT DATE 6. LER NUMBER 7. REPORT DATE 8. OTHER FACILITIES INVOLVED SEQUENTIAL I Rev FACti tTY NAME UUvl\t:l NUMBER MONTH DAY YEAR YEAR NUMBER NO. MONTH DAY YEAR NA NA FACILITY NAME DOCKET NUMBER 03 15 2016 2016-001 -00 05 12 2016 NA NA 9. OPERATING
: 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) MODE D 20.2201(b)
D 20.2203(a)(3)(i)
D 50.73(a)(2)(i)(C)
D 50.73(a)(2)(vii)
D 20.2201(d)
D 20.2203(a)(3)(ii)
IZl 50.73(a)(2)(ii)(A)
D 50. 73(a)(2)(viii)(A)
MODE5 D D D D 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B)
D 20.2203(a)(2)(i)
D 50.36(c)(1)(i)(A)
D 50. 73(a)(2)(iii)
D 50. 73(a)(2)(ix)(A)
: 10. POWER LEVEL D 20.2203(a)(2)(ii)
D 50.36(c)(1
)(ii)(A) D 50.73(a)(2)(iv)(A)
D 50.73(a)(2)(x)
D 20.2203(a)(2)(iii)
D 50.36(c)(2)
D 50.73(a)(2)(v)(A)
D 73.71 (a)(4) D 20.2203(a)(2)(iv)
D 50.46(a)(3)(ii)
D 50.73(a)(2)(v)(B)
D 73.71 (a)(5) 0% D D D D 20.2203(a)(2)(v) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(C)
OTHER D 20.2203(a)(2)(vi)
D 50.73(a)(2)(i)(B)
D 50.73(a)(2)(v)(D) below or in 12.
CONTACT FOR "HIS LER FACILITY NAME TELEPHONE NUMBER (Include Area Code) Thomas P. Schneider, Senior Licensing Engineer 920-755-7797
: 13. COMPLETE ONE LINE FOR EACH = 'I= LUI DESCRIBED IN THS REPORT CAUSE SYSTEM COMPONENT MANU-REPORTABLE CAUSE SYSTEM COMPONENT MANU-REPORTABLE FACTURER TO EPIX FACTURER TO EPIX 8 C8 v C635 y NA NA NA NA NA
: 15. EXPECTED MONTH DAY YEAR SUBMISSSION YES (If yes, complete 15. EXPECTED SUBMISSION DATE) X NO DATE NA NA NA ABSTRACT (Limit to 1400 spaces i.e, approximately 15 single-spaced typewritten lines) At 0649 on March 15, 2016 with Unit 1 shut down in MODE 5 for refueling activities, a boric acid indication upstream of the valve seating surface on the inlet of the valve body of 1 CV-2008, Letdown Orifice 8 Outlet Control Valve was identified as a through-wall flaw. The flaw location was within the reactor coolant system (RCS) pressure boundary as defined by 10 CFR 50.2, "Definitions." The valve body is original plant equipment.
This event is being reported pursuant to 10 CFR 50. 73(a)(2)(ii)(A) for material defects in the primary coolant system that were not acceptable in accordance with ASME Section XI.
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017 (02-2014)
LICENSEE EVENT REPORT (LER) CONTINUATION SHEET 1. FACILITY NAME 2. DOCKET Point Beach Nuclear Plant Unit 1 05000266 NARRATIVE Description of the Event: Estimated burden per response to comply with this mandatory collection request: 80 hours. Reported lessons learned are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate to the FOIA, Privacy and lnfonmation Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555*0001,or byintemet e-mail to lnfocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-1 0202, (3150-01 04), Office of Management and Budget, Washington, DC 20503. If a means used to impose an infonmation collection does not display a currently valid OMS control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the infonmation collection.
: 6. LERNUMER YEAR I SEQUENTIAL I NUMBER 2016 -001 REV NO. 00 2 3. PAGE OF 2 At 0649 on March 15, 2016 with Unit 1 shut down in MODE 5 for refueling activities, a boric acid indication upstream of the valve seating surface on the inlet of the valve body of 1 CV-2008, Letdown Orifice B Outlet Control Valve was identified as a through-wall flaw. The flaw location was within the reactor coolant system (RCS) pressure boundary as defined by 10 CFR 50.2, "Definitions." The valve body is original plant equipment.
The Unit 1 Letdown Orifice B Outlet Control Valve (1 CV-2008) has been replaced and returned to service. This 60 day licensee event report is being submitted in accordance with the requirements of 10 CFR 50. 73(a)(2)(ii)(A) for any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being degraded.
The material defect in the primary coolant system could not be found acceptable in accordance with ASME Section XI, IWB-3600, "Analytical Evaluation of Flaws" or ASME Section XI, Table IWB-3410-1, "Acceptance Standards." Cause of the Event: The most likely cause of the degraded barrier was localized corrosion near a sand inclusion indication in the valve body. Analysis of the Event: Maintenance activities identified a through-wall flaw on the valve body of the Letdown Orifice B Outlet Control Valve. The flaw location was upstream of the valve seating surface, on the inlet of the valve body. 1 CV-2008 is one of three valves in parallel in the same application.
The valve was within the RCS pressure boundary, which is defined as being connected to the RCS, up to and including the outermost containment isolation valve in system piping which penetrates the primary reactor containment.
A causal evaluation determined the through-wall flaw in the valve body to most likely be the result of localized corrosion near a sand inclusion indication.
The indication was identified during the manufacturing examination and testing processes.
The indication was accepted by the manufacturer using the acceptance criteria of ASTM Specification E71, "Reference Radiographs for Steel Casting up to 2 inches in Thickness." The localized corrosion most likely allowed system contents to leech through the sand inclusion.
The valve body is original plant equipment and has been in service the entire life of the plant. The flaw has been corrected by replacement of the valve body. Corrective Actions: The Unit 1 Letdown Orifice B Outlet Control Valve (1 CV-2008) has been replaced and returned to service. Destructive testing will be performed on the removed valve to validate the apparent cause. Additional actions/reviews will be taken as required following receipt of the destructive testing report. Examinations have been completed as part of the extent of condition on three valves from the same heat number that had manufacturing indications.
A corrective action has been created to perform a visual examination of one additional valve during the next refueling outage. Safety Significance:
The event was determined to be of very low safety significance.
1 CV-2008 can be isolated from the RCS by a remotely operated valve 1 RC-427, Reactor Coolant Loop B Cold Leg to Charging and Volume Control System (CVCS) Letdown Isolation Valve.
Components with similar sand inclusion indications have been or will be visually examined for evidence of boric acid indications.
While the condition was present, there was no loss of any safety systems, structures or components needed to shut down the reactor, maintain safe shutdown conditions, remove residual heat, control the release of radioactive material or mitigate the consequences of an accident.
There was no impact on the health and safety of the public as a result of this condition.
Similar Events: There have not been similar events of this degraded condition in the past three years. Component Failure Data: 2 INCH-1500 LB COPES-VULCAN MODEL 63 GLOBE VALVE}}

Latest revision as of 18:44, 12 July 2018

Redirect to:

Retrieved from "https://kanterella.com/w/index.php?title=ML16133A170&oldid=880782"