ML16076A187
| ML16076A187 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 03/11/2016 |
| From: | Schwarz C Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML16076A187 (6) | |
Text
Tennessee Valley Authority, Post Office Box 2000, Soddy Daisy, Tennessee 37384-2000 March 11, 2016 10 CFR 50.4 10 CFR 50.55a ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuciear Plant, Unit 2 Renewed Facility Operating License No. DPR-79 NRC Docket No. 50-328
Subject:
Unit 2 Cycle 20 Refueling Outage Day Inservice Inspection Summary Report In accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Xl, Article IWA-6230 and Code Case N-532-5, the Tennessee Valley Authority is providing the Sequoyah Nuclear Plant (SQN), Unit 2, Inservice Inspection (ISI) Summary Report within 90 days from completion of the Unit 2 Refueling Outage 20 (U2R20). The U2R20 refueling outage ended on December 15, 2015. Accordingly, this report is required to be submitted by March 14, 2016. The report contains the Owner's Activity Report, a list of items with flaws or relevant conditions that required evaluation for continued service (Table 1), an abstract of repair and replacement activities required for continued service due to a flaw or relevant condition (Table 2), and the evaluation of acceptability of inaccessible areas of the steel containment vessel in accordance with 10 CFR 50.55a(b)(2)(ix)(A).
There are no regulatory commitments associated with this submittal. Should you have any questions, please contact Michael McBrearty at (423) 843-7170.
Respectfully, Site Vice President Sequoyah Nuclear Plant
Enclosure:
ASME Section XI Inservice Inspection Summary Report, Unit 2 Cycle 20 Refueling Outage cc (Enclosure):
NRC Regional Administrator - Region II
.(,
NRC Senior Resident Inspector - Sequoyah Nuclear Plant NRC Project Manager - Sequoyah Nuclear Plant k
ENCLOSURE TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNIT 2 ASME SECTION XI INSERVICE INSPECTION
SUMMARY
REPORT UNIT 2 CYCLE 20 REFUELING OUTAGE
FORM OAR-I OWNER'S ACTIVITY REPORT Report Number U2R20 Piant Unit No SEQUOYAH NUCLEAR PLANT. P.O. Box 2000. Soddy-Daisy. TN 37384-2000 O.
Commercial service date June 1, 1982 Refueling outage no.
20 2
(if applicable)
Current inspection interval Current inspection period Third Inspection Inerval
)1~ 2~, 3'~ 4' other)
Third Inspection Period Edition and Addenda of Section XI applicable to the inspection plans 2001 Edition, 2003 Addenda Date and revision of inspection plans November 23, 2015- 0-SI-DXI-000-114.3, Revision 28 Edition and Addenda of Section XI applicable to repair/replacement activities, if different than the inspection plans_____
Not Applicable Code Cases used:
N-460, N-513-3, N-532-4. N-532-5, N-566-2, N-586-1, N-686-1, N-706-1, N-716-1, N-722-1, N-729-1, N-770-1, N-798, N-800 (if applicable, including cases modified by Case N-532 and later revisions)
CERTIFICATE OF CONFORMANCE I certify that (a) the statements made in this report are correct; (b) the examinations and tests meet the Inspection Plan required by the ASME Code Xl and (c) the repair/replacement activities and evaluations supporting the completion of U2C20 conform to the requirements of SectiOn Xl.
(refueling outage number)
Signed Date Z3/*h,,
f / *Owr~4yrOwner'sDesignee, *fle
)
CERTIFICATE OF INSERVICE INSPECTION I, the undersigned, holding a valid commission issued by the National Board of Boiler and Pressure Vessel Inspectors and employed by The Hartford Steam Boiler Inspection and Insurance Company of Connecticut have inspected the items described in this Owner's Activity Report, and state that, to the best of my knowledge and belief, the Owner has performed all activities represented by this report in accordance with the requirements of Section Xl.
By signing this certificate neither the Inspector nor his employer makes any warranty, expressed or implied, concerning the repair/replacement activities and evaluation described in this report. Furthermore, neither the Inspector nor his employer shall be liable in m esonal injury or property damage or a loss of any kind arising from or connected with this inspection.
____-____Commissions
/3 7/2--
AA/*'/ 2--
Inspector's Signature National Board, State, Province, and Endorsements ASME Code Section Xl Exception-
Reference:
TVA Condition Report 1098901,
Subject:
EPRI NDE Alert 201 3-09 for bolting exams and NRC Enforcement Guidance Memorandum (EGM) 14-003 enforcement discretion for the use of generic EPRI bolt and stud NDE procedures when doing ASME Section Xl bolting exams.
E-1
ASME SECTION Xl INSERVICE INSPECTION
SUMMARY
REPORT SEQUOYAH NUCLEAR PLANT UNIT 2 CYCLE 20 REFUELING OUTAGE TABLE 1 ITEMS WITH FLAWS OR RELEVANT CONDITIONS THAT REQUIRED EVALUATION FOR CONTINUED SERVICE Examination Category and Item Description Evaluation Description Item Number N-566-2 Evaluation, Assuming RC Loop 3 Letdown Flow valve, SQN maximum probably corrosion B-P B1.10 FCV-062-0069-A, Leak with small amounts damage shows negligible wear, B-,B1.0 of dry discolored boron on the body to affected studs and nuts have no bonnet bolting.
visible corrosion damage, WO
______________________________117381236 retightened.
Visual inspection indicated none of Safety Injection System, CCP Flow Element, teblswr xoe olaae C-H, C7. 10 SQN-2-FE-063-0029, small amount of dry Bligi tils te.La a
white boric acid found on the orifice plate stopped under WO 117380277.
Seal Water isolation test connection, SQN None of the 8 studs were exposed to C-H, C7.10 VLV-062-0715, dry white boron was found leakage. Bolting is stainless steel.
between the body and end cap connections.
Leak is not active.
F-A, F1.20C Component ID 2-MSH-342, Variable Spring Design basis spring setting tolerance Support, improper setting on support, both is 10% in accordance with SQN-DC-spring hangers out of range (NOI 2-SQ-445)
V-24.2. The as-found spring setting remains acceptable and meets design basis requirements.
The as found setting was outside the Component ID 2-MSH-344, Variable Spring design tolerance of 10%. Work Order F-A, F1.20C Support, improper setting found on support.
117379624 reset the spring in (NOI 2-SQ-443) accordance with the drawing.
Successive exam performed.
The loose jam nut is inconsequential since the opposite jam nut remains adequate to prevent rotation. This does not render the tie-rod non-Component ID 2-CRDH-1, CRDM Seismic functional and is acceptable. Cotter F-A, Fl1.40 Support, Loose jam nuts and 3 cotter pins pins, while loose, remain functional in were easily removed, (NOI 2-SQ-441) preventing clevis pins from dislodging. This condition does not render tie-rods non-functional and is also acceptable. Procedure 0-MI-MRR-068-000.0 corrected these conditions.
Maximum anchor load calculation F-A, Fl1.40 Component ID RVH-1, Reactor Vessel determined that as found loose nuts support, 3 loose nuts (NOI 2-SQ-442) are acceptable and both RV supports remain fully functional.
E-2
ASME SECTION XI INSERVICE INSPECTION
SUMMARY
REPORT SEQUOYAH NUCLEAR PLANT UNIT 2 CYCLE 20 REFUELING OUTAGE TABLE 2 ABSTRACT OF REPAIRS, REPLACEMENTS, OR CORRECTIVE MEASURES REQUIRED FOR CONTINUED SERVICE Codel Item D
escription Date IRepairRelcementf Class Description of Work Completed Plan Number No Applicable Repair/Re lacement activities occurred durnn this period.
E-3
ASME SECTION Xl INSERVICE INSPECTION
SUMMARY
REPORT SEQUOYAH NUCLEAR PLANT UNIT 2 CYCLE 20 REFUELING OUTAGE REPORTING REQUIRED BY 10 CFR 50.55a(b)(2)(ix)(A)
ASME Section XI, Subsection IWE Steel Containment Vessel Inspection Program 10 CFR 50.55a(b)(2)(ix)(A) requires reporting of the degradation assessment for inaccessible areas when conditions are identified in accessible areas during the performance of the ASME Section Xl, Subsection IWE Steel Containment Vessel (SCy) Inspection Program that could indicate the presence of or result in degradation to such inaccessible areas.
The conditions identified during the Unit 2 Cycle 20 refueling outage on the SCV in an accessible area were determined to be acceptable and do not affect the degradation assessment for inaccessible areas.
E-4
Tennessee Valley Authority, Post Office Box 2000, Soddy Daisy, Tennessee 37384-2000 March 11, 2016 10 CFR 50.4 10 CFR 50.55a ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuciear Plant, Unit 2 Renewed Facility Operating License No. DPR-79 NRC Docket No. 50-328
Subject:
Unit 2 Cycle 20 Refueling Outage Day Inservice Inspection Summary Report In accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Xl, Article IWA-6230 and Code Case N-532-5, the Tennessee Valley Authority is providing the Sequoyah Nuclear Plant (SQN), Unit 2, Inservice Inspection (ISI) Summary Report within 90 days from completion of the Unit 2 Refueling Outage 20 (U2R20). The U2R20 refueling outage ended on December 15, 2015. Accordingly, this report is required to be submitted by March 14, 2016. The report contains the Owner's Activity Report, a list of items with flaws or relevant conditions that required evaluation for continued service (Table 1), an abstract of repair and replacement activities required for continued service due to a flaw or relevant condition (Table 2), and the evaluation of acceptability of inaccessible areas of the steel containment vessel in accordance with 10 CFR 50.55a(b)(2)(ix)(A).
There are no regulatory commitments associated with this submittal. Should you have any questions, please contact Michael McBrearty at (423) 843-7170.
Respectfully, Site Vice President Sequoyah Nuclear Plant
Enclosure:
ASME Section XI Inservice Inspection Summary Report, Unit 2 Cycle 20 Refueling Outage cc (Enclosure):
NRC Regional Administrator - Region II
.(,
NRC Senior Resident Inspector - Sequoyah Nuclear Plant NRC Project Manager - Sequoyah Nuclear Plant k
ENCLOSURE TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNIT 2 ASME SECTION XI INSERVICE INSPECTION
SUMMARY
REPORT UNIT 2 CYCLE 20 REFUELING OUTAGE
FORM OAR-I OWNER'S ACTIVITY REPORT Report Number U2R20 Piant Unit No SEQUOYAH NUCLEAR PLANT. P.O. Box 2000. Soddy-Daisy. TN 37384-2000 O.
Commercial service date June 1, 1982 Refueling outage no.
20 2
(if applicable)
Current inspection interval Current inspection period Third Inspection Inerval
)1~ 2~, 3'~ 4' other)
Third Inspection Period Edition and Addenda of Section XI applicable to the inspection plans 2001 Edition, 2003 Addenda Date and revision of inspection plans November 23, 2015- 0-SI-DXI-000-114.3, Revision 28 Edition and Addenda of Section XI applicable to repair/replacement activities, if different than the inspection plans_____
Not Applicable Code Cases used:
N-460, N-513-3, N-532-4. N-532-5, N-566-2, N-586-1, N-686-1, N-706-1, N-716-1, N-722-1, N-729-1, N-770-1, N-798, N-800 (if applicable, including cases modified by Case N-532 and later revisions)
CERTIFICATE OF CONFORMANCE I certify that (a) the statements made in this report are correct; (b) the examinations and tests meet the Inspection Plan required by the ASME Code Xl and (c) the repair/replacement activities and evaluations supporting the completion of U2C20 conform to the requirements of SectiOn Xl.
(refueling outage number)
Signed Date Z3/*h,,
f / *Owr~4yrOwner'sDesignee, *fle
)
CERTIFICATE OF INSERVICE INSPECTION I, the undersigned, holding a valid commission issued by the National Board of Boiler and Pressure Vessel Inspectors and employed by The Hartford Steam Boiler Inspection and Insurance Company of Connecticut have inspected the items described in this Owner's Activity Report, and state that, to the best of my knowledge and belief, the Owner has performed all activities represented by this report in accordance with the requirements of Section Xl.
By signing this certificate neither the Inspector nor his employer makes any warranty, expressed or implied, concerning the repair/replacement activities and evaluation described in this report. Furthermore, neither the Inspector nor his employer shall be liable in m esonal injury or property damage or a loss of any kind arising from or connected with this inspection.
____-____Commissions
/3 7/2--
AA/*'/ 2--
Inspector's Signature National Board, State, Province, and Endorsements ASME Code Section Xl Exception-
Reference:
TVA Condition Report 1098901,
Subject:
EPRI NDE Alert 201 3-09 for bolting exams and NRC Enforcement Guidance Memorandum (EGM) 14-003 enforcement discretion for the use of generic EPRI bolt and stud NDE procedures when doing ASME Section Xl bolting exams.
E-1
ASME SECTION Xl INSERVICE INSPECTION
SUMMARY
REPORT SEQUOYAH NUCLEAR PLANT UNIT 2 CYCLE 20 REFUELING OUTAGE TABLE 1 ITEMS WITH FLAWS OR RELEVANT CONDITIONS THAT REQUIRED EVALUATION FOR CONTINUED SERVICE Examination Category and Item Description Evaluation Description Item Number N-566-2 Evaluation, Assuming RC Loop 3 Letdown Flow valve, SQN maximum probably corrosion B-P B1.10 FCV-062-0069-A, Leak with small amounts damage shows negligible wear, B-,B1.0 of dry discolored boron on the body to affected studs and nuts have no bonnet bolting.
visible corrosion damage, WO
______________________________117381236 retightened.
Visual inspection indicated none of Safety Injection System, CCP Flow Element, teblswr xoe olaae C-H, C7. 10 SQN-2-FE-063-0029, small amount of dry Bligi tils te.La a
white boric acid found on the orifice plate stopped under WO 117380277.
Seal Water isolation test connection, SQN None of the 8 studs were exposed to C-H, C7.10 VLV-062-0715, dry white boron was found leakage. Bolting is stainless steel.
between the body and end cap connections.
Leak is not active.
F-A, F1.20C Component ID 2-MSH-342, Variable Spring Design basis spring setting tolerance Support, improper setting on support, both is 10% in accordance with SQN-DC-spring hangers out of range (NOI 2-SQ-445)
V-24.2. The as-found spring setting remains acceptable and meets design basis requirements.
The as found setting was outside the Component ID 2-MSH-344, Variable Spring design tolerance of 10%. Work Order F-A, F1.20C Support, improper setting found on support.
117379624 reset the spring in (NOI 2-SQ-443) accordance with the drawing.
Successive exam performed.
The loose jam nut is inconsequential since the opposite jam nut remains adequate to prevent rotation. This does not render the tie-rod non-Component ID 2-CRDH-1, CRDM Seismic functional and is acceptable. Cotter F-A, Fl1.40 Support, Loose jam nuts and 3 cotter pins pins, while loose, remain functional in were easily removed, (NOI 2-SQ-441) preventing clevis pins from dislodging. This condition does not render tie-rods non-functional and is also acceptable. Procedure 0-MI-MRR-068-000.0 corrected these conditions.
Maximum anchor load calculation F-A, Fl1.40 Component ID RVH-1, Reactor Vessel determined that as found loose nuts support, 3 loose nuts (NOI 2-SQ-442) are acceptable and both RV supports remain fully functional.
E-2
ASME SECTION XI INSERVICE INSPECTION
SUMMARY
REPORT SEQUOYAH NUCLEAR PLANT UNIT 2 CYCLE 20 REFUELING OUTAGE TABLE 2 ABSTRACT OF REPAIRS, REPLACEMENTS, OR CORRECTIVE MEASURES REQUIRED FOR CONTINUED SERVICE Codel Item D
escription Date IRepairRelcementf Class Description of Work Completed Plan Number No Applicable Repair/Re lacement activities occurred durnn this period.
E-3
ASME SECTION Xl INSERVICE INSPECTION
SUMMARY
REPORT SEQUOYAH NUCLEAR PLANT UNIT 2 CYCLE 20 REFUELING OUTAGE REPORTING REQUIRED BY 10 CFR 50.55a(b)(2)(ix)(A)
ASME Section XI, Subsection IWE Steel Containment Vessel Inspection Program 10 CFR 50.55a(b)(2)(ix)(A) requires reporting of the degradation assessment for inaccessible areas when conditions are identified in accessible areas during the performance of the ASME Section Xl, Subsection IWE Steel Containment Vessel (SCy) Inspection Program that could indicate the presence of or result in degradation to such inaccessible areas.
The conditions identified during the Unit 2 Cycle 20 refueling outage on the SCV in an accessible area were determined to be acceptable and do not affect the degradation assessment for inaccessible areas.
E-4