IR 05000313/2017007
| ML17265A274 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 09/21/2017 |
| From: | Thomas Farnholtz Region 4 Engineering Branch 1 |
| To: | Richard Anderson Entergy Operations |
| References | |
| IR 2017007 | |
| Download: ML17265A274 (19) | |
Text
September 21, 2017
Richard Anderson, Site Vice President
Arkansas Nuclear One Entergy Operations, Inc.
1448 SR 333 Russellville, AR 72802-0967
SUBJECT: ARKANSAS NUCLEAR ONE, UNITS 1 AND 2 - NRC DESIGN BASES ASSURANCE INSPECTION (PROGRAMS) REPORT 05000313/2017007 AND
Dear Mr. Anderson:
On August 10, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection
at your Arkansas Nuclear One, Units 1 and 2.
The NRC inspectors discuss ed the results of this inspection with Mr. T. Evans, Acting Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report.
NRC inspectors documented one finding of very low safety significance (Green) in this report. This finding involved a violation of NRC requirements.
If you contest the violation or significance of the NCV, 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 IV; the Director, Office of Enforcement; and the NRC resident inspector at the Arkansas Nuclear One.
If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement 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 IV; and the NRC resident inspector at Arkansas Nuclear One. 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 10 CFR 2.390, "Public Inspections, Exemptions, Requests for Withholding.
"
Sincerely,/RA/
Thomas R. Farnholtz, Chief Engineering Branch 1
Division of Reactor Safety Docket Nos. 50-313 and 50-368 License Nos. DPR-51 and NPF-6
Enclosure:
Inspection Report 05000313/2017007 and 05000368/2017007
w/Attachment:
Supplemental Information
cc: Electronic Distribution
ML17265A274 SUNSI Review ADAMS: Non-Publicly Available Non-Sensitive Keyword: By: JDrake Yes No Publicly Available Sensitive NRC-002 OFFICE RI:EB1 SRI:EB2 SRI:EB2 C:PBE C:EB1 NAME J. Braisted S. Alferink J. Drake N. O'Keefe T. Farnholtz SIGNATURE /RA/ /RA/ /RA/ /RA/ /RA/ DATE 9/11/17 9/20/17 9/20/17 9/21/17 9/21/17 Enclosure U.S. NUCLEAR REGULATORY COMMISSION REGION IV Docket: 05000313 and 05000368 License: DPR-51 and NPF-6 Report: 05000313/2017007 and 05000368/2017007 Licensee: Entergy Operations, Inc.
Facility: Arkansas Nuclear One, Units 1 and 2 Location: Junction of Highway 64 West and Highway 333 South Russellville, Arkansas Dates: July 24 through August 10, 2017 Team Leader: J. Drake, Senior Reactor Inspector, Engineering Branch 2 Inspectors: S. Alferink, Reactor Inspector J. Braisted, Reactor Inspector Approved By: Thomas R. Farnholtz, Branch Chief Engineering Branch 1
Division of Reactor Safety
2
SUMMARY
IR 05000313/2017007; 05000368/2017007 07/24/2017 - 08/10/2017; Arkansas Nuclear One, Units 1 and 2; Inspection Procedure 71111.21N, Design Bases Assurance (Programs)
The inspection activities described in this report were performed between July 24, 2017, and August 10, 2017, by three inspectors from the NRC's Region IV office. One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. The significance of inspection findings is indicated by their color (Green, White, Yellow, or Red), which is determined using Inspection Manual Chapter 0609, "Significance Determination Process" dated April 29, 2015. Cross-cutting aspects are determined using Inspection Manual Chapter 0310, "Aspects Within the Cross-Cutting Areas" dated December 4, 2014. Violations of NRC requirements are dispositioned in accordance with the NRC's Enforcement Policy dated November 1, 2016. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process."
A. NRC Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green.
The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance are promptly identified and corrected. Specifically, from 1996 until August 10, 2017, the licensee failed to properly resolve the environmental conditions in room 38 following a high-energy line break, even when challenged during a self-assessment by members of the quality assurance group in June 29, 2015. In response to this issue, the licensee determined that in the event of a break in the letdown line, an engineered safety feature signal automatic closure of both the inside and outside reactor building isolation valves occurs in approximately 40 seconds, prev enting room 38 from going "harsh." This finding was entered into the licensee's corrective action program as Condition Report CR-ANO-1-2017-02441.
The inspectors determined that the licensee's failure to adequately evaluate and take prompt corrective actions to resolve an identified condition adverse to quality related to the high energy line break analysis for room 38 was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, when the licensee identified that the environmental conditions in room 38 of the auxiliary building were "harsh," as determined by Design Bases Calculation CALC-01-EQ-1002-02, they failed to properly resolve the condition adverse to quality. In accordance with Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At-Power," dated June 19, 2012, Exhibit 2, "Mitigating Systems Screening Questions," dated July 1, 2012, the inspectors determined that the finding was of very low safety significance (Green) because the finding (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with the licensee's maintenance rule program. The finding has a cross-cutting aspect in the area of human resources, training, because the organization failed to provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, there was a lack of understanding of the current licensing bases for the plant displayed by engineering, operations, and management [H.9]. (Section 1R21.2.2)
B. Licensee Identified Violations
None 4
REPORT DETAILS
REACTOR SAFETY
1R21 Design Basis Assurance Inspection (Programs)
a. Inspection Scope
The inspectors performed an inspection as outlined in NRC Inspection Procedure 71111.21N, Attachment 1, "Environmental Qualification under 10 CFR 50.49, Programs, Processes, and Procedures." The inspectors assessed the implementation of the environmental qualification program as required by 10 CFR 50.49, "Environmental qualification of electric equipment important to safety for nuclear power plants," by Arkansas Nuclear One, Units 1 and 2. The inspectors evaluated whether Arkansas Nuclear One, Units 1 and 2, staff properly maintained the environmental qualification of electrical equipment important to safety throughout plant life, established and maintained required environmental qualification documentation records, and implemented an effective corrective action program to identify and correct environmental qualification-related deficiencies.
The inspection included a review of environmental qualification program procedures, component environmental qualification files, environmental qualification test records, equipment maintenance and operating history, maintenance and operating procedures,
vendor documents, design documents, and calculations. The inspectors interviewed program owners, engineers, maintenance staff, and warehouse staff. The inspectors performed in-plant walkdowns (where accessible) to verify equipment was installed as described in the environmental qualification component documentation files for Arkansas Nuclear One, Units 1 and 2, and that the components were installed in their tested configuration. Additionally, the inspectors performed in-plant walkdowns to determine whether equipment surrounding the components could fail in a manner that could prevent the safety functions of the components and to verify that components located in areas susceptible to a high-energy line break were properly evaluated for operation in a "harsh" environment. The inspectors reviewed and inspected the storage of replacement parts and associated procurement records to verify environmental qualification parts approved for installation in the plant were properly identified and controlled, and that storage and environmental conditions did not adversely affect the components' qualified lives. Documents reviewed for this inspection are listed in the attachment.
In accordance with the inspection procedure, the inspectors initially selected 10 components to assess the adequacy of the environmental qualification program. Component samples selected for this inspection were:
- CV-1407, Unit 1, Borated Water Storage Tank T-3 Outlet Valve Operator
- SV-3840, Unit 1, CV-3840 Control Air Isolation Solenoid Valve for Decay Heat Pump P-34A Cooling Jacket Service Water Inlet Valve
- ZS-2613, Unit 1, Internal Switch for CV-2613 Valve Operator for Steam to Emergency Feedwater Pump Turbine P-7A
- P-7B, Unit 1, Emergency Feedwater Pump Motor
- 2SV-1010-1A, Unit 2, Solenoid Air Control Valve for Main Steam Isolation Valve 2CV-1010-1
- 2CV-1025-1, Unit 2, Internal Switch for 2CV-1025-1 Valve Operator For Emergency Feedwater Pump 2P-7B discharge valve to Steam Generator 2E24A
- 2LT-1031-1, Unit 2, Steam Generator 2E-24A Downcomer Level Transmitter
- 2PT-4601-1, Unit 2, Reactor Coolant System Pressurizer Pressure Transmitter
- 2VSF-1C, Unit 2, Containment Cooler Fan Motor
- 2VSF-1A, Unit 2, Containment Cooler Fan Motor During the course of the inspection, the following item was included in the scope:
- Electric governor for P-7A, Unit 1, Emergency Feedwater Pump Turbine
b. Findings
Failure to Promptly Identify and Correct an Inadequate Design Bases Calculation
Introduction.
The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," associated with the licensee's failure to adequately evaluate and take prompt corrective actions to address an identified condition adverse to quality related to the reactor coolant system letdown line break analysis for room 38 of the auxiliary building.
Description.
While reviewing Calculation NES-13, "Environmental Qualification (EQ) - Environmental Service Conditions," Revision 13, the inspectors noted that the environment in room 38 (emergency feedwater pump room) following a letdown line break was considered "harsh," with the temperature in the room exceeding 150 degrees Fahrenheit (°F) as determined by Design Bases Calculation CALC-01-EQ-1002-01, "Reactor Coolant Letdown Line HELB Analysis," Revision 2. Title 10 CFR 50.49, "Environmental Qualification of Electric Equi pment Important to Safety for Nuclear Power Plants," requires the licensee to place safety-related electric equipment that is relied upon to remain functional during and following design basis events in the environmental qualification program. Neither of the emergency feedwater pumps or any of their subcomponents were within the scope of the environmental qualification program, even though both of them were safety-related components.
Reviewing the licensee's environmental qualification program, the inspectors noted that both trains of the emergency feedwater system had temperature-sensitive components. Revision 2 of CALC-01-EQ-1002-01 indicated that the design basis letdown line break would result in the temperature of the room housing the emergency feedwater pumps (room 38) exceeding 150 °F, making it a "harsh" environment. Since a 6 letdown line break could result in actuation of both emergency feedwater pumps, the ability of both the motor-driven and the turbine-driven pumps to perform their intended function could be challenged during this event.
The inspectors informed the licensee of their concern on August 7, 2017.
During follow-up discussions, the inspectors determined that the licensee had previously entered this issue into their corrective action program as Condition Report CR-ANO-1-2015-02630 in March 2015. Subsequent review of this condition report revealed that the condition report was closed with the resolution that, "EFW (emergency feedwater) is not essential in this case since cooldown may be accomplished by continuance of normal main feedwater flow from either MFW pump in conjunction with the condensate system."
The inspectors reviewed the following information:
- Calculation ER-93-R-1040-01 indicates that the emergency feedwater motor-driven pump motor bearings are qualified to a room ambient temperature of 148
°F.
- Calculation ER-93-R-1040-01 indicates that the Woodward governor on the emergency feedwater turbine driven pump is qualified to a room ambient temperature of 150
°F.
- Design Basis Calculation CALC-01-EQ-1002-01 states that for the design bases event of a reactor coolant system letdown line break, the temperature in room 38 exceeds 150
°F and both trains of emergency feedwater would be lost.
- Safety Analysis Report, Section 4.3.6, "Reliance On Interconnected Systems," states that the turbine or electric driven emergency feedwater pump supply feedwater to the steam generators and the resultant steam is vented to the atmosphere to provide required cooling during design bases events.
- Safety Analysis Report, Section 14.2.2.5.2.2, "ECCS Performance Analysis Assumptions," states that the turbine generator and main feedwater pumps are assumed to trip coincident with the reactor trip.
- The assumptions in Calculation ER-93-R-1040-01 state that the design break in the reactor coolant letdown line will result in a net loss from the reactor coolant system of 48 pounds mass per second in excess of the reactor coolant system make up system capacity and result in a drop in pressurizer level of approximately 24 inches per minute until the leak is isolated.
Additionally, the emergency feedwater motor-driven pump motor bearings and the
Woodward governor on the emergency feedwater turbine-driven pump were previously in the environmental qualification program. They were removed from the program in 1986 based on modifications to the upper south penetration room (room 77) that resulted in establishment of a new high-energy line break vent path for the main feedwater line break. As a result of the modifications, a main feedwater line break would exhaust into the boiler room from room 77 and would no longer affect room 38. At that time, the licensee determined that room 38 would be considered a mild environment.
7 In 1996, the licensee reevaluated the reactor coolant letdown line high energy line break and determined that room 38 would be a "harsh" environment based on the assumptions
used in Design Bases Calculation CALC-01-EQ-1002-01, Revision 2. When this was identified, rather than place the emergency feedwater motor-driven pump motor bearings and the Woodward governor on the emergency feedwater turbine-driven pump back in the environment qualification program, the licensee made the determination that emergency feedwater was not required for this design bases event and they could rely on main feedwater and condensate to provide required cooling. This is the same justification that was used to close Condition Report CR-ANO-1-2015-02630.
The licensee did not initiate a condition report on the NRC's concern, even after the inspectors questioned if that was in accordance with their corrective action procedures.
The licensee stated that the issue did not represent an immediate safety concern because they had previously performed operability assessments for the affected areas, which established a reasonable expectation for operability pending resolution of the identified issue. After the inspectors presented the licensee with the information in the safety analysis report regarding the turbine generator and main feedwater pumps, the licensee reevaluated the assumptions used in CALC-01-EQ-1002-01, Revision 2, and determined that the assumed time to isolate the leak was overly conservative and a more appropriate time to isolate the leak was five minutes. The licensee determined that in the event of a break in the letdown line, an engineered safeguards signal automatic closure of both the inside and outside reactor building isolation valves occurs in approximately 30 to 40 seconds. The Channel 2 engineered safety feature actuation signal (engineered safety feature actuation signal) (low RCS pressure or high reactor building pressure) will also initiate closure of CV-1221 (Aux Building), and engineered safety feature actuation signal Channel 1 (low RCS pressure or high reactor building pressure) will initiate closure of CV-1214 and CV-1216 (Letdown coolers outlet Reactor building side) isolating the break flow and preventing room 38 from going "harsh". The licensee determined that securing the leak within five minutes would prevent the environment in room 38 from becoming "harsh" for the reactor coolant letdown line break. The licensee entered the issue into the corrective action program on August 9, 2017, as Condition Report CR-ANO-1-2017-02441. Because the assumptions in the design bases calculation were determined by the licensee to be overly conservative, the
equipment would not have been subjected to a "harsh" environment, function would not have been lost.
The inspectors determined that the licensee had failed to promptly correct a condition adverse to quality. The original acceptance by engineering, operations, and management that it was acceptable to rely on main feedwater to provide required cooling during design bases events points to a training deficiency associated with understanding the current licensing bases throughout the organization.
Analysis.
The inspectors determined that the licensee's failure to adequately evaluate and take prompt corrective actions to resolve an identified condition adverse to quality related to the high energy line break analysis for room 38 was a performance deficiency.
The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, when the licensee identified that the environmental conditions in room 38 of the auxiliary building were "harsh," as determined by Design Bases Calculation CALC-01-EQ-1002-02, they 8 failed to properly resolve the condition adverse to quality. In accordance with Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At-Power," dated June 19, 2012, Exhibit 2, "Mitigating Systems Screening Questions," the inspectors determined that the finding was of very low safety significance (Green) because the finding (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with the licensee's maintenance rule program. The finding has a cross-cutting aspect in the area of human resources, training, because the organization failed to provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, there was a lack of understanding of the current licensing bases for the plant displayed by engineering, operations, and management [H.9].
Enforcement.
The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance are promptly identified and corrected. Contrary to the above, from 1996, until August 10, 2017, the licensee failed to establish measures to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance were promptly identified and corrected. Specifically, the licensee failed to properly resolve the environmental conditions in room 38 following a high-energy line break, even when challenged during a self-assessment by members of the quality assurance group in June 29, 2015. In response to this issue, the licensee determined that in the event of a break in the letdown line, an engineered safety feature signal automatic closure of both the inside and outside reactor building isolation valves occurs in approximately 40 seconds, prev enting room 38 from going "harsh". This finding was entered into the licensee's corrective action program as condition report CR-ANO-1-2017-02441. Because this finding was of very low safety significance and has been entered into the licensee's corrective action program, this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000313/2017007-01, "Failure to Promptly Identify and Correct an Inadequate
Design Bases Calculation."
OTHER ACTIVITIES
4OA2 Problem Identification and Resolution
The inspectors reviewed condition reports associated with the selected components, operator actions, and operating experience notifications. There were issues with the resolution of a significant percentage of the condition reports reviewed by the inspectors. However, because of the limited number of condition reports reviewed and the narrow focus of condition reports associated with the 10 components selected for the inspection, this is not a statistically significant evaluation of the licensee's overall corrective action program. The inspectors noted that the licensee had difficulty tracking and trending 9 items related to the environmental qualification program in their corrective action program based on the fact that condition reports known to the inspectors that had issues impacting or related to the environmental qualification program were not provided in the list of condition reports initially provided by the licensee.
4OA6 Meetings, Including Exit
Exit Meeting Summary
On August 10, 2017, the inspectors presented the inspection results to Mr. T. Evans, Acting Vice President, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
- T. Evans, Vice President, Coordinator
- R. Putnam, Director, Chief Engineer
- V. Bacanskas, Director, Chief Engineer
- P. Butler, Manager, Design and Program Engineering
- J. Kirkpatrick, General Manager, Plant Operations
- B. Lynch, Manager, Radiation Protection
- S. Morris, Manager, Chemistry
- R. Penfield, Director, Regulatory Assurance
- G. Sullins, Senior Manager, Recovery
- S. Pyle, Manager, Regulatory Assurance
- M. Skartvedt, Manager, System Engineering
- D. Vogt, Senior Manager, Operations
- N. Mosher, Licensing Specialist, Regulatory Assurance
NRC Personnel
- T. Farnholtz, Branch Chief, Engineering Branch 1
- C. Henderson, Senior Resident
- T. Sullivan, Resident Inspector
- M. Tobin, Resident Inspector
- J. Choate, Project Engineer
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
- 05000313/2017007-01 NCV Failure to Promptly Identify and Correct an Inadequate Design Bases Calculation (Section 1R21.2.2)
LIST OF DOCUMENTS REVIEWED
Calculations
- Number Title Revision 98-R-0008-01 Determination of Service Temperatures for EQ
- Rosemount Transmitter Qualified Lives
- CALC-00-R-0003-
- CALC-88-EQ-
- CALC-93-EQ-
- 0002-01 ASCO Solenoid Valve Qualified Life 4 CALC-ANO1-EQ-
- 11-00001 Cycling Screening of the
- ANO-1 EQ Equipment to Evaluate Extension of Qualification into the License Renewal Period
- CALC-04-EQ-
- 0001-01 Qualified Life of Reliance Motors with RH and RN
- Insulation Class
- CALC-11-E-0001-
- Calculation of Qualified Lives of Various Components in Support of
- ANO-1 License Renewal
- CALC-90-EQ-
- 2002-04 Analysis of EQ Equipment Installed Outside Containment at Ambient >105F
- CALC-93-EQ-
- 0002-01 ASCO Solenoid Valve Qualified Life 4
- CALC-93-R-1026-
- Replacement & PUPR
- CALC-99-EQ-
- 0001-09
- ANO-2 40-Year Normal Dose Inside the Reactor Building
- CALC-ANO1-EQ-
- 11-00001 Cycling Screening of the
- ANO-1 EQ Equipment to Evaluate Extension of Qualification into the License Renewal Period
- CALC-01-EQ-
- 1002-01 Reactor Coolant Letdown Line HELB Analysis 2
- Design Change Packages Number Title Revision
- Update EQ Documents Based on Latest Rosemount Test Report Revs
- CR-ANO-2-2013-00282, CA-002
- EC-63812 Incorporate
- CR-ANO-C-2016-00884 Op-Eval on
- CV-1000 Orientation; NRC Follow-Up Questions and Responses into ANO EQ Files; and Any Additional Information from Extent of Condition Review
Procedures
- Number Title Revision 1203.045 Rapid Plant Shutdown 18 1203.039 Excess RCS Leakage 16
- 25.006 Environmentally Qualified Equipment Maintenance Program 24 1202.001 Reactor Trip 37 1202.002 Loss Of Subcooling Margin 10
- 202.005 Inadequate Core Cooling 10
- 203.026 Loss Of Reactor Coolant Makeup 14
- 1412.001 Preventive Maintenance of Limitorque SB/SMB Motor Operators
- 1413.039 ASCO 8320 DC Watertight Solenoid Enclosure
(Nuclear)
- 1601.101 Operation of High Pressure Water Decontamination Systems 3
- EN-FAP-LI-001 Performance Improvement Review Group (PRG)
- Process 10
Procedures
- Number Title Revision
- EN-LI-102 Corrective Action Program 29
- EN-MA-141 Limitorque Valve Operator Model SMB/SB/SBD-000 Through 5 MOV and HBC Periodic Inspection
Condition Reports
(Reviewed)
- CR-ANO-1-2017-02441
- CR-ANO-1-2001-00243
- CR-ANO-1-2017-00711
- CR-ANO-1-2015-02630
- CR-ANO-2-2017-03422
- CR-ANO-2-2017-03243
- CR-ANO-2-2017-03718
- CR-ANO-2-2017-03789
- CR-ANO-2-2017-03998
Condition Reports
- Generated During the Inspection
- CR-ANO-1-2017-02158
- CR-ANO-1-2017-02159
- CR-ANO-1-2017-02160
- CR-ANO-1-2017-02297
- CR-ANO-1-2017-02299
- CR-ANO-2-2017-04294
- CR-ANO-2-2017-04297
- CR-ANO-2-2017-04298
- CR-ANO-2-2017-04299
- CR-ANO-2-2017-04301
- CR-ANO-2-2017-04341
- CR-ANO-2-2017-04452
- CR-ANO-2-2017-04477
- CR-ANO-2-2017-04488
- CR-ANO-C-2017-02726
- CR-ANO-C-2017-02862
- CR-ANO-C-2017-03010
Work Orders
- 00435564
- 00469259
- 00477931
- 00478873
- 00437211
Miscellaneous
- Number Title Revision Date
- NES-01 Environmental Qualification (EQ) Program 5
- NES-13 Environmental Qualification - Environmental Service Conditions
- BQR-88-0228 Discontinuation of Rebuild Kits for ASCO "NP" Series Valves May 23, 1989 LO-ALO-2012-
- 00091 Snapshot Assessment: EQ Program January 31, 2015
Miscellaneous
- Number Title Revision Date LO-ALO-2015-
- 00074 Snapshot Assessment in Recovery to One:
- Program March 31, 2016 LO-ALO-2016-
- 00087 Environmental Qualification (EQ) Program Assessment June 25, 2017 87-EQ-0003-05 Typical MOV Thermal Lag Analysis Following a Main Steam Line Break
- ER-002800 EQ Approval of Rosemount Reports D8300040 Rev. E and D8400102 Rev. F
- ASCO Field Notifications December 5, 1989
- Clarification of Information Related to the Environmental Qualification of Limitorque Motorized Valve Operators August 1989 2A087 2VSF-1C 9 B044
- SV-3840 4
- B089
- CV-1407,
- ZS-1407 9
- B199
- CV-2613,
- ZS-2613 8
- Environmental Qualification Test Reports Number Title Revision Date V06-001 Solenoid Valve 6 V33-002 Motorized Valve Actuator - AC OC 9
- V41-002 Electric Motor 7
- 600456 Limitorque Report December 19, 1975
- 600461 Limitorque Report June 7, 1976
- Environmental Qualification Test Reports Number Title Revision Date
- AQR-67368 ASCO Report 1 AQS21678/TR ASCO Report A
- B0009 Limitorque Report April 30, 1976
- B0058 Limitorque Report January 11, 1980
- V06-032 Test Report No.
- AQR-67368/Rev. 1 Report on Qualification of Automatic Switch Co. (ASCO) Catalog
- NP-1 Solenoid Valves for Safety-Related Applications in Nuclear Power Generating Stations November 2, 1988 V06-036 Investigation into Eliminating the Use of Lubricants on the NP Series of Valves March 11, 2006 V33-064 Clarification of Information Related to the Environmental Qualification of Limitorque Motorized Valve Operators September 13, 1989V33-077 Comparative Analysis of Nebula and MOV Long Life Greases for Limitorque Main Gearbox
- Applications November 4, 2002 V43-048 Qualification Report for Pressure Transmitter Model 1154 Rosemount Repot D8400102
- October 25, 1984 V43-090 Qualification Report for Pressure Transmitters Rosemount Model 1153 Series D Rosemount
- Repot D8300040 Revision A August 26, 1985 V43-130 Qualification Report for Rosemount Model 1154 Series H Pressure Transmitter Rosemount Report
- D8700096 Revision D June 26, 1989 V43-167 Qualification Report for Rosemount Model 1154 Series H Pressure Transmitter Rosemount Report
- 08700096 Revision K April 4, 2012 V43-169 IEEE Qualification Report:
- Span Thermistor Replacement for 1153 Series B, 1153 Series D,
- 1154 and 1154 Series H Pressure Transmitters Rosemount Nuclear Instruments, Inc. Document Number
- 02011020 Revision A May 22, 2012
- Environmental Qualification Data Packages Number Title Revision 2A087 2VSF-1C 5 B044
- SV-3840 1
- B089
- CV-1407,
- ZS-1407 4
- B199
- CV-2613,
- ZS-2613 6