IR 05000327/2020010
| ML20112F443 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 04/21/2020 |
| From: | James Baptist NRC/RGN-II/DRS/EB1 |
| To: | Jim Barstow Tennessee Valley Authority |
| References | |
| IR 2020010 | |
| Download: ML20112F443 (18) | |
Text
April 21, 2020
SUBJECT:
SEQUOYAH NUCLEAR PLANT, UNITS 1 & 2 - DESIGN BASIS ASSURANCE INSPECTION (PROGRAMS) INSPECTION REPORT 05000327/2020010 AND 05000328/2020010
Dear Mr. Barstow:
On March 13, 2020, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Sequoyah Nuclear Plant, Units 1 & 2. On April 17, 2020, the NRC inspectors discussed the results of this inspection with Mr. Matthew Rasmussen and other members of your staff. The results of this inspection are documented in the enclosed report.
One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violation or the significance or severity of the violation documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement; and the NRC Resident Inspector at Sequoyah Nuclear Plant, Units 1 & 2.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; and the NRC Resident Inspector at Sequoyah Nuclear Plant, Units 1 & 2. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely,
/RA/
James B. Baptist, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos. 05000327 and 05000328 License Nos. DPR-77 and DPR-79
Enclosure:
Inspection Report
Inspection Report
Docket Numbers: 05000327 and 05000328 License Numbers: DPR-77 and DPR-79 Report Numbers: 05000327/2020010 and 05000328/2020010 Enterprise Identifier: I-2020-010-0053 Licensee: Tennessee Valley Authority Facility: Sequoyah Nuclear Plant, Units 1 & 2 Location: Soddy Daisy, TN Inspection Dates: February 24, 2020 to March 13, 2020 Inspectors: M. Greenleaf, Reactor Inspector G. Ottenberg, Senior Reactor Inspector R. Patterson, Senior Reactor Inspector A. Rosebrook, Senior Reactor Analyst Approved By: James B. Baptist, Chief Engineering Branch 1 Division of Reactor Safety Enclosure
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a design basis assurance inspection (programs) inspection at Sequoyah Nuclear Plant, Units 1 & 2, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
Inadequate Guidance for Motor Operated Valve Capability Assumptions Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [P.3] - 71111.21N.
NCV 05000327,05000328/2020010-01 Resolution 02 Open/Closed The inspectors identified a Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, subsection 55a(b)(3)(ii), for the licensees failure to establish a program that ensured that all the design basis safety functions of certain motor-operated valves (MOVs) would be met. Specifically, since at least identification of the issue in 2016, the licensee had not incorporated appropriate guidance for ensuring that the leak tight safety function of inservice test (IST) Category A valves would be met into their MOV program documents.
Additional Tracking Items
None.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.21N.02 - Design-Basis Capability of Power-Operated Valves Under 10 CFR 50.55a Requirements POV Review (IP Section 03)
The inspectors:
a. Determined whether the sampled POVs are being tested and maintained in accordance with NRC regulations along with the licensees commitments and/or licensing bases.
Specific Guidance b. Determined whether the sampled POVs are capable of performing their design-basis functions.
c. Determined whether testing of the sampled POVs is adequate to demonstrate the capability of the POVs to perform their safety functions under design-basis conditions.
d. Evaluated maintenance activities including a walkdown of the sampled POVs (if accessible).
- (1) 1-FCV-67-146, Component Cooling Heat Exchanger 1A1/1A2 Discharge Control Valve
- (2) 2-FCV-63-73, Containment Sump Flow Isolation Valve
- (5) 1-FSV-68-394, Reactor Vessel Head Vent Isolation Valve
- (6) 2-LCV-3-148, Steam Generator Level Control Valve
- (7) 1-FCV-68-333, RCS Pressurizer Relief Flow Control Valve
- (8) 1-FCV-63-001, Refueling Water Storage Tank to RHR Pump Flow Control Valve
INSPECTION RESULTS
Inadequate Guidance for Motor Operated Valve Capability Assumptions Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [P.3] - 71111.21N.0 NCV 05000327,05000328/2020010-01 Resolution 2 Open/Closed The inspectors identified a Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, subsection 55a(b)(3)(ii), for the licensees failure to establish a program that ensured that all the design basis safety functions of certain motor-operated valves (MOVs) would be met. Specifically, since at least identification of the issue in 2016, the licensee had not incorporated appropriate guidance for ensuring that the leak tight safety function of inservice test (IST) Category A valves would be met into their MOV program documents.
Description:
During the inspection, the inspectors observed that the licensees calculation design standard, DS-M18.2.21, "Motor Operated Valve Thrust and Torque Calculations," Revision 25, referenced General Engineering Specification G-50, "Torque and Limit Switch Settings for Motor-Operated Valves," Revision 11, and stated: Torque switch controlled MOVs that set the bypass limit switch at >98% (Reference General Engineering Specification G-50) can base margin on design/limit switch controlled equation. The inspectors were concerned that the licensee could misapply the calculation of actuator thrust capability to motor-operated valves that are classified as Category A within the licensees IST program. Valves considered Category A are valves that have a leak tight function. The concern was that the design/limit switch-controlled equation allows full actuator capability to be used for comparison against the valve required thrust in their margin calculations, rather than a lower, more limiting value based on the torque switch setting. The full actuator capability is appropriate to be assumed to be applied until the design/limit switch controlled circuit de-energizes the MOV. However, it is inappropriate to be assumed for MOV program margin calculation purposes if a valve is demonstrated to perform its leak tight safety function after setting the design/limit switch controlled circuit under static conditions and expect it to perform under design basis differential pressure (dP) and flow conditions.
In Generic Letter (GL) 89-10, "Safety-Related Motor-Operated Valve Testing and Surveillance," the NRC staff requested licensees to verify the design-basis capability of their safety-related MOVs and to establish long-term motor-operated valve (MOV) programs. On December 21, 1989, TVA responded to the GL, and made the following commitment: "TVA will develop and implement a comprehensive motor-operated valve testing and surveillance program for Browns Ferry, Watts Bar, and Sequoyah Nuclear Plants satisfying the intent of Generic Letter 89-10 by June 28, 1994." On September 14, 1990, the NRC responded to TVA's commitment letter with the following: "In your response to GL 89-10 dated December 21, 1989, you committed to develop and implement a program to satisfy the intent of GL 89-10. The staff interprets this as a commitment to meet the schedule and recommendations provided in the generic letter and its supplement dated June 13, 1990."
Supplement 1 to GL 89-10, "Results of the Public Workshops," dated June 13, 1990, contained the following position in question 32 regarding the demonstration necessary to show that valve safety functions in the closed position were met:
"May an MOV be considered closed simply if the torque switch trips in the last 2 to 3 percent of the stroke?
NRC staff response - No. The staff considers an MOV to be closed if the applicable system train is capable of meeting all of its required safety functions with the MOV in that position, and the MOV meets the necessary leak-tightness criteria. A licensee cannot assume that an MOV will be adequately closed simply because the torque switch was bypassed until the last 2 to 3 percent of the stroke. In terms of the generic letter, the staff will accept closure of an MOV to within 2 to 3 percent during a test to have demonstrated that the MOV is operable in the close direction under design-basis conditions, provided the MOV will perform its safety function in that position. As this would leave almost no operability margin, the licensee would need to demonstrate that the test fully met the design-basis conditions, with degraded voltage."
In 2016, the licensee reviewed their MOV program, as documented in Document No. 3640, ASME [American Society of Mechanical Engineers] OM [Operation and Maintenance] Code Appendix III Gap Assessment for TVAs Nuclear Stations, Revision 1, and determined that General Engineering Specification G-50 should be updated. Comments in the gap analysis stated that the licensee needed to, provide basis for the 98% close torque switch bypass to ensure valve closure, define how to calculate margin, and if it is permissible to use for valves that have specific leakage requirements.
The licensee generated a condition report (CR 1271906) and updated the G-50 specification (Rev. 11) in January 2020 to include additional guidance for isolation valves that are considered IST Category A. The guidance stated:
For safety-related, GL89-10/App. III Program MOVs which are required for containment isolation/LLRT [local leak rate test], control switch trip must be set at or beyond hard seat contact (C11). This may be accomplished by making the valve position or limit seated, or by setting the torque switch so that the valve can produce the required thrust to achieve hard seat contact under maximum dP conditions."
The inspectors concluded the updated guidance was insufficient because the licensee did not demonstrate the new control switch trip setting (at or beyond hard seat contact) would ensure that the leak tight function would be met. The licensee assumed that additional loading that would occur after the control switch trip would provide sufficient sealing load under design basis conditions. However, the licensee did not have an analysis or test data to support their judgement.
Inspectors determined that the licensee misapplied the calculation of actuator capability to 13 valves within the program. Ten of the affected valves required re-evaluation for operability because they were calculated to have negative margin when considering torque switch control instead of limit switch control. The station wrote CR 1590551 to address the inspectors concern and determined that the 13 valves with a leak tight requirement that utilize this setup assumption would be able to provide their safety functions. For some of the affected valves, the licensee changed their assumptions by crediting measured inertial loading or reanalyzed thrust requirements to demonstrate margin between the torque switch setting and the required thrust during the applicable containment isolation event. The inspectors noted that although the licensee had plans to revise all Sequoyah GL 89-10 / 96-05 MOV calculations as documented in CR 1459470, the updates would have been based on the recently revised G-50 guidance which was insufficient, without additional justification, to demonstrate the leak tight functions would be met.
Corrective Actions: After the inspectors identified the concern, the licensee reviewed the 13 affected valves that had bypass limit switches set to >98% and determined that they could demonstrate the valves would remain operable. The licensee also generated CR 1602141 to address vulnerabilities in the MOV program guidance documents.
Corrective Action References: CRs 1602141 and 1590551.
Performance Assessment:
Performance Deficiency: The licensees failure to incorporate appropriate guidance into fleet design standard DS-M18.2.21 and fleet General Engineering Specification G-50 to ensure that the leak tight safety function of IST Category A valves would be met was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the SSC and Barrier Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, Inspection Manual Chapter 0612, Appendix E, More than minor example 3a states: When revising the calculation, the licensee had to do one of the following:
- (a) use a different calculation methodology because the original methodology resulted in unfavorable margin;
- (b) revise assumptions solely to obtain favorable results; (c)revise other calculations in order to establish operability or functionality; or
- (d) the remaining margin falls outside the licensees design process acceptance criteria. Unfavorable margin means that had the correct values been used originally, the licensees design process would not have accepted the modification. In order to provide reasonable assurance that the safety function would be met, the licensee had to revise their assumptions for some of the affected valves to obtain favorable results.
Significance: The inspectors assessed the significance of the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Exhibit 3, "Barrier Integrity," the finding screened as having very low safety significance (Green) because the finding did not represent a an actual open pathway in the physical integrity of reactor containment; did not represent the failure of containment pressure control equipment; did not represent the failure of containment heat removal components; and did not involve the actual reduction in function of hydrogen igniters in the reactor containment.
Cross-Cutting Aspect: P.3 - Resolution: The organization takes effective corrective actions to address issues in a timely manner commensurate with their safety significance. Following identification of the issue in 2016, the licensee completed corrective actions by updating the MOV program guidance in January 2020, but the actions taken were not effective in that the MOV program after correction did not ensure all the valve safety functions would be met.
Enforcement:
Violation: Title 10 of the Code of Federal Regulations (10 CFR), subsection 50.55a(b)(3)(ii)required, in part, that licensees must establish a program to ensure that MOVs continue to be capable of performing their design basis safety functions. Contrary to this, since at least identification of the issue in 2016, the licensee did not establish a program that ensured the valves that had a leak tight function were capable of performing all their design basis safety functions. Specifically, the licensee had not correctly incorporated guidance for demonstrating that the leak tight safety function of containment isolation valves would be met into their MOV program guidance.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On April 17, 2020, the inspectors presented the design basis assurance inspection (programs) inspection results to Mr. Matthew Rasmussen and other members of the licensee staff.
- On March 12, 2020, the inspectors presented the onsite debrief inspection results to Mr.
Matthew Rasmussen and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection Type Designation Description or Title Revision or
Procedure Date
71111.21N.02 Calculations 00D53EPMRJP061091 Generic Letter 89-10 MOV Population at Sequoyah Revs. 19
Units 1 & 2 and 20
03D53EPMWLL063094 AFW Hydraulic Analysis Rev. 16
1-FCV-63-001 Documentation of Design Basis Review, Required Rev. 5
Thrust Calc and Valve & Actuator Capability
Assessment for 1-FCV-63-001
1-FCV-63-008 Documentation of Design Basis Review, Required Rev. 2
Thrust Calc and Valve & Actuator Capability
Assessment for 1-FCV-63-008
1-FCV-63-011 Documentation of Design Basis Review, Required Rev. 4
Thrust Calc and Valve & Actuator Capability
Assessment for 1-FCV-63-011
1-FCV-67-146 DOCUMENTATION OF DESIGN BASIS REVIEW, Rev. 3
REQUIRED TORQUE CALC AND VALVE AND
ACTUATOR CAPABILITY
ASSESSMENT FOR 1-FCV-67-146
1-FCV-68-333 Documentation of Design Basis Review, Required Rev. 7
Thrust Calc and Valve & Actuator Capability
Assessment for 1-FCV-68-333
2-FCV-63-073 DOCUMENTATION OF DESIGN BASIS REVIEW, Rev. 5
REQUIRED THRUST CALC AND VALVE &
ACTUATOR
CAPABILITY ASSESSMENT FOR 2-FCV-63-073
2-FCV-70-134 Documentation of Design Basis Review, Required Rev. 3
Thrust Calc and Valve and Actuator Capability
Assessment for 2-FCV-70-134
MD000099920040148 Setpoint Review for Category 2 AOVs Rev. 10
MDQ00000320020129 AFW AOV System Review Rev. 1
MDQ00000320020130 Evaluation of Required Thrust for MDAFW Large Rev. 1
Inspection Type Designation Description or Title Revision or
Procedure Date
MDQ00006720000095 ERCW Flow Balance Hydraulic Model Rev. 17
MDQ0000682017000375 Reactor Head Vent Capacity Analysis Rev. N/A
MDQ00099920080190 Documentation of GL 89-10 Henry Pratt Butterfly Rev. 4
Valve Static Testing Acceptance Criteria
MDQ00099920110249 Sequoyah Nuclear Plant JOG MOV Periodic Rev. 0
Verification Classification
MDQ0009992014000158 Analysis of DP test data to support the SON MOV Rev. 2
Program
MDQ0999980039 MOV Thrust Requirements Considering Pressure Rev. 5
Lock Using ComEd Methodology
OSG70020 Safety Injection System (063) 10 CFR 50.49 Rev. 19
Category and Operating Times
SCG-4M-00794 Seismic Analysis Crane Aloyco 12 lnch 300 LB Rev. 4
Valves
SQNETAPAC Auxiliary Power System Revs. 95
and 96
SQS20200 SQN Probabilistic Risk Assessment -MOV Risk Rev. 3
Ranking
Corrective 1017597
Action 1018105
Documents 1108722
1157371
211795
237177
241396
271906
297216
1316892
1338800
1365730
1366082
1395498
1405334
Inspection Type Designation Description or Title Revision or
Procedure Date
1459470
1494559
1497075
1498362
1567574
1570428
1570436
1570440
1570523
1580811
1586882
1589446
1592228
2691
753504
SQ971023PER
SQPER921503
Corrective CR 1589744 SQN POV DBAI 2020010 Clean Boron from stem,
Documents CR 1589753 SQN POV DBAI 2020010 unrestrained ladder in U1
Resulting from 669 PC along right side wall 20 ft in Czone
Inspection CR 1589781 SQN POV DBAI 2020010 - Walkdown of ERCW
Mezzanine HK Issues
CR 1589999 SQN POV DBAI 2020010 - Incorrect Units within
CR 1590230 SQN POV DBAI 2020010 - Incorporation of EPRI
Guide 3002008055 into DS-M18.2.21
CR 1590489 SQN POV DBAI 2020010 - Incorrect EQ
Temperature used in GL 89-10 Calculation
CR 1590513 SQN POV DBAI 2020010 - Update CAT Record
CR 1590551 SQN POV DBAI 2020010 - Category A MOVs
Identified Which Take Credit for 98% Torque Switch
Inspection Type Designation Description or Title Revision or
Procedure Date
Bypass
CR 1590580 SQN POV DBAI 2020010 - DCA Not Incorporated
into WO 119465059
CR 1591403 SQN POV DBAI 2020010
CR 1591559 SQN POV DBAI 2020010 - EQ Binder status
CR 1592321 SQN POV DBAI 2020010 - excessive grease found
on actuator for 1-FCV-70-8
CR 1592546 SQN POV DBAI 2020010 - Incorrect Stem Factor 03/05/2020
used in GL 89-10 Calculations 1- and 2-FCV-63-008
CR 1593643 SQN POV DBAI 2020010 - Discrepancy between
GL 89-10 calculation 2-FCV-63-073 and
00D53EPMRJP061091
CR 1593694 NRC POV DBAI 2020010 - Lack of critical thinking
for not including inertial loading into MOV Calc
CR 1593729 SQN POV DBAI 2020010 - Clarify 110% Extended
Actuator Use in DS-M18.2.21
CR 1602141 SQN POV DBAI 2020010 - Lack of Guidance for
MOV Margin Determination for All Accide
Drawings 02-403-0244 Limitorque Valve Control Diagram Rev. 9
1, 2-45N779-26 Wiring Diagram 480V Shutdown Aux Power Rev. 33
Schematic Diagrams SH 26
1, 2-45N779-48 Wiring Diagrams 480V Shutdown Aux Power Rev. 4
Schematic Diagrams SH-48
1,2- 47W432-2 Mechanical Residual Heat Removal System Piping Rev. 11
1,2-47W813-1 Flow Diagram Reactor Cooling System Rev. 58
1,2-47W845-2 Mechanical Flow Diagram- Essential Raw Cooling Rev. 115
Water System
1-47A941-104 Thrust Requirements for 1-FCV-68-333 Rev. 1
1-47A941-166 Torque Requirements for Motor Operated Butterfly Rev. 2
Valves
1-47A941-40 Thrust Requirements for Motor Operated Valve 1- Rev. 1
FCV-63-008
1-47W811-1 Flow Diagram Safety Injection System Rev. 77
Inspection Type Designation Description or Title Revision or
Procedure Date
2-47A941-48 Thrust Requirements for Motor Operated Valve 2- Rev. 0
FCV-63-073
2-47W432-1 Mechanical Residual Heat Removal System Piping Rev. 6
2-47W803-2 Flow Diagram Auxiliary Feed Water Rev. 78
2-47W811-1 Flow Diagram Safety Injection System Rev. 62
79AB-001 Solenoid Oper Globe Valve High Temp, High Press Rev. M
Energize to Open 1" SW
88415 Velan 3 Inch Forged Bolted Bonnet Valve Rev. 8
94-13295 ASA Series 300 8" No. S70W DD Weld Ends Rev. 8
Outside Screw & Yoke Double Disc Gate Valve with
SMB-1 Limitorque Valve Control, Lip Seal and Limit
Switches
94-13300 Double Disc Gate Valve Stainless Steel, Weld Ends, Rev. 1
Outside Screw w/ SMB-3 Limitorque Act. & Limit
Switches Size: 18" Class (S70) 300
C-8152 General Arrangement HB Operator SMB000 Motor Rev. 901
E-2303 Cross Section ASME Section III Standard Nuclear dated
Class 3 N-MKII Valve 5/10/72
K-7634-E Crane 300lb Cast Alloy Split-Wedge Disc Valve Rev. 3
Engineering D22542 Resolve Component Cooling System Concerns Rev. A
Changes Related to Appendix R
D22564 Modify Motor Operated Valves (MOVs) to meet JOG Rev. A
Class A or B Requirements
DCN 21550 AFW Level Control Valves-Actuator Capability Rev. A
Increase
DCN 22501 Increase the capability of 2-FCV-63-72 & -73 to Rev. A
operate under differential pressure
DCN 23623 Degraded Non-Conforming MOV Modifications - Rev. A
Gears/Actuators/Motor/Cable Reroute
DCN S-13223 Values shown on drawing series 47A941 for Rev. A
"Minimum Thrust at Torque Switch Trip" do not
agree with the values shown in the supporting
calculations. Revise drawings as required.
Inspection Type Designation Description or Title Revision or
Procedure Date
DCN X00156 Modify Dwgs of Sump Valve Room Manways Rev. B
EDC E-20067 Revise 47A941 series drawings to incorporate new Rev. A
switch settings
EQV 23070 Replace Motor SQN-2-MTRB-063-0073-B Rev. A
SQN-100516 MOV Stem Thread Removal Option to Facilitate Rev. 0
Future QSS Installation for Selected Valves
Engineering 2761C Sequoyah Nuclear Plant JOG MOV Periodic Rev. 0
Evaluations Verification Classification
EWR-15-PEG-063-579 Approve Drawing and Design Reports for ASME 07/02/2015
Valve Disc; TVA P.O. #768647; Flowserve's SO #
110676; Typ TIIC # CVP637B and CVQ672F
RIMS B88940823002 Engineering Analysis of Torque Requirements and dated
Material, Design and Dimensional Data for Nuclear 7/15/94
Butterfly Valves
SCG4M00766 Seismic Qualification of Anchor Darling 18" 300lb Rev. 7
Double Disc Gate Valve (Motor Operated)
SCG4M00816 Seismic Qualification of 24" Henry Pratt Butterfly Rev. 3
Valves, UNID Nos. 1-FCV-67-146, 2-FCV-67-146,
and 0-FCV-67-152 (Mk No. 47W450-20)
SQNEQ-EM-028 Essentially Mild Documentation for Limitorque Rev. 13
Actuators
SQS40077 RHR Sump Valve Room - Safety Evaluation dated
5/20/82
Miscellaneous DS-M18.2.21 Motor Operated Valve Thrust and Torque Rev. 25
Calculations
Equipment Specification Motor Operated Valves for TVA Sequoyah Nuclear Rev. 1
678765 Plants Units 1 and 2
ER-5.0 Equipment Inaccuracy for Motor Operated Valves Rev. 27
G-50 Torque, Thrust and Control Switch Settings for Revs. 10
Motor-Operated Valves and 11
G-955186 Solenoid Operated Globe and Throttle Valves Rev. 1
ASME Boiler and Pressure Vessel Code Section III
Class 1, 2, and 3
ML020360077 COMMENTS ON JOINT OWNERS' GROUP AIR dated
Inspection Type Designation Description or Title Revision or
Procedure Date
OPERATED VALVE PROGRAM DOCUMENT 10/8/99
NPG-SPP-09.26.13 Air Operated Valve Program Rev. 0
NPG-SPP-09.26.14 Motor Operated Valve Program Rev. 4
RIMS B38920730807 Limitorque Corporation Letter- Actuator Data dated
Analysis 7/29/92
SQN-DC-V-21.0 Environmental Design Rev. 28
SQN-DC-V-27.3 Sequoyah Nuclear Plant- Safety Injection System Rev. 24
SQN-DC-V-27.3 Sequoyah Nuclear Plant - Safety Injection System Rev. 24
SQN-DC-V-27.4 Sequoyah Nuclear Plant - Reactor Coolant System Rev. 25
SQN-DC-V-7.4 Essential Raw Cooling Water System (67) Rev. 34
SQNEQ-MOV-005 Limitorque Actuators Outside Containment With Rev. 47
Brakes
SQNEQ-SOL-010 Target Rock Solenoid Valves Rev. 23
Procedures 0-SI-SVX-063-266.0 ASME Valve Code Testing Rev. 0035
0-SI-SXV-063-266.0 ASME Code Valve Testing Rev. 34
dated
11/17/18
0-SI-SXV-067-266.0 ASME Code Valve Testing Rev. 35
dated
10/16/19
0-SI-SXV-068-266.0 ASME Code Valve Testing Rev. 14
0-TI-SXI-000-200.0 Inservice Testing Program Rev. 4
2-SI-SLT-088-156.0 Containment Integrated Leak Rate Test Rev. 2
dated
2/24/06
2-SO-63-5 Emergency Core Cooling System Rev. 62
NPG-SPP-22.300 Corrective Action Program Rev. 17
Self- CR 1563644 Self Assessment for Power Operated Valves Rev. 0
Assessments (POVs) in regards to NRC Inspection Procedure 71111.21N.02, "Design-Bases Capability of Power-
Operated Valves Under 10 CFR 50.55a
Requirements"
Work Orders 111980783
Inspection Type Designation Description or Title Revision or
Procedure Date
113806733
115439910
115682058
115682538
116284999
116510557
117165093
117732973
118147355
118614256
118800994
119245809
119446769
119941824
19446720
19446769
15