IR 05000338/2019010
| ML19225D175 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 08/13/2019 |
| From: | James Baptist NRC/RGN-II |
| To: | Stoddard D Virginia Electric & Power Co (VEPCO) |
| References | |
| IR 2019010 | |
| Download: ML19225D175 (22) | |
Text
ust 13, 2019
SUBJECT:
REISSUE - NORTH ANNA POWER STATION, UNITS 1 AND 2 - DESIGN BASIS ASSURANCE INSPECTION (PROGRAMS) INSPECTION REPORT 05000338/2019010 AND 05000339/2019010
Dear Mr. Stoddard:
On July 18, 2019, the US Nuclear Regulatory Commission (NRC) issued the subject inspection report for the North Anna Power Station, Units 1 and 2, Adams Accession Number ML19199A178. In reviewing this report, we identified administrative discrepancies throughout the report. Therefore, we are reissuing a revised version of Inspection Report 05000338/2019010 and 05000339/2019010. We request that you disregard the previous information and replace the previous cover letter and report with this report in its entirety.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at https://www.nrc.gov/reading-rm/adams.html. (The Public Electronic Reading).
Room). I apologize for any inconvenience this error may have caused. If you have any questions, please contact me at (404) 997-4506.
Sincerely,
/RA/
James B. Baptist, Chief Engineering Branch 1 Docket Nos.: 05000338 and 05000339 License Nos: NPF-4 and NPF-7
Enclosure:
1. North Anna Power Station, Units 1 and 2 Design Basis Assurance Inspection (Programs)
Cover Letter 2. North Anna Power Station, Units 1 and 2 Design Basis Assurance Inspection (Programs)
IR 05000338/2019010 and 05000339/2019010
Inspection Report
Docket Numbers: 05000338 and 05000339 License Numbers: NPF-4 and NPF-7 Report Numbers: 05000338/2019010 and 05000339/2019010 Enterprise Identifier: I-2019-010-0024 Licensee: Virginia Electric and Power Company (VEPCO)
Facility: North Anna Power Station, Units 1 and 2 Location: Mineral, Virginia 23117 Inspection Dates: May 13, 2019 to June 07, 2019 Inspectors: B. Collins, Reactor Inspector T. Fanelli, Senior Reactor Inspector M. Greenleaf, Reactor Inspector T. Su, Reactor Inspector Approved By: James B. Baptist, Chief Engineering Branch 1 Division of Reactor Safety Enclosure 2
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a design basis assurance inspection (teams) inspection at North Anna Power Station, Units 1 and 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 Failure to Qualify Containment Components for Initial Acidic Chemical Environment Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green None (NPP) 71111.21N Systems NCV 05000338,05000339/2019010-
Closed The inspectors identified a Green finding and associated Non-cited Violation (NCV) of 10 CFR 50.49(e)(3) when the licensee failed to qualify all 10 CFR 50.49 electrical equipment for the most severe chemical effects expected in containment.
Failure to Verify the Adequacy of Design for Aging and Accident Profile Testing Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green None (NPP) 71111.21N Systems NCV 05000338,05000339/2019010-
Closed The inspectors identified three examples of a Green finding and associated Non-cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III when the licensee failed to verify compliance with environmental qualification (EQ) requirements in accordance with design control manual NDCM-3.3 Rev. 11, Design Verification.
Additional Tracking Items Type Issue Number Title Report Section Status URI 05000338,05000339/2 Potential 50.9 Violation for 71111.21N Open 019010-03 Incomplete Information Submittal
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 - Design Bases Assurance Inspection (Programs)
The inspectors evaluated the environmental qualification program implementation through the sampling of the following components: (IP Section 02.01)
Primary Containment (Inside Containment) ===
- (1) Victoreen Radiation Monitoring System
- (2) Gamma-Metrics Neutron Flux Monitors
- (3) Boston Insulated Wire (BIW) - Cable - Cross Linked Polyethylene (XLPE) / Neoprene Jacket
- (4) Pressurizer Vent Line Solenoid Operated Valve 2-RC-SOV-202A1 (Target Rock)
- (5) Connectron Terminal Block NSS3
- (6) PORV Limit Switch 1-RC-PCV-1455C (ZS-PCV-1455C-A1) (NAMCO)
(Outside Containment) (2 Samples)
- (2) Validyne Engineering Corporation, Multiplexer, Cables, Connectors, Inadequate Core Cooling System
INSPECTION RESULTS
Failure to Qualify Containment Components for Initial Acidic Chemical Environment Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green None (NPP) 71111.21N Systems NCV 05000338,05000339/2019010-01 Closed The inspectors identified a Green finding and associated Non-cited Violation (NCV) of 10 CFR 50.49(e)(3) when the licensee failed to qualify all 10 CFR 50.49 electrical equipment for the most severe chemical effects expected in containment.
Description:
During design basis accidents in which equipment important to safety must perform their safety function, the quench spray (QS) system actuates to reduce pressure and temperature in containment. The QS system is designed to take water from an internal weir inside the refueling water storage tank (RWST) and pump it to spray rings inside containment. The sprayed water reduces the temperature and pressure associated with the design basis accident. The sprayed water from the QS system also provides a certain amount of "iodine scrubbing" to the containment atmosphere in which postulated radioactive iodine would be captured by the water as it fell, entrapping it in solution. To accomplish this, the water is buffered with sodium hydroxide (NaOH) to a basic pH (pH > 7.0).
The design of the QS system includes a 5-minute time delay for sodium hydroxide injection into the RWST. By licensee calculations and UFSAR Section 6.2.3.1.1, the initial spray pH in containment prior to the mixing of sodium hydroxide in the spray could be as low as 4.7. The licensees environmental zone description describes the chemical spray as 8.6-10.6 pH from 0-4 hours then 7-8.6 from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to termination of sprays.
The result of this discrepancy is that the licensee has never qualified any equipment inside containment for this initial chemical spray condition in which only diluted boric acid is being sprayed into containment. The initial period of spray could be significant because, even though the duration of exposure to the acidic conditions would be relatively short, the equipment in containment would absorb (be wetted) and potentially retain the acidic solution throughout the accident.
An inspection sample, BIW 300 V instrument cable, was reviewed by the inspectors to determine if any chemical incompatibility existed between the materials of construction and the diluted boric acid. Using various chemical compatibility tables, the inspectors determined that the neoprene jacket could be adversely affected by the effects of the boric acid spray.
Corrective Actions: The licensee gathered additional information from other nuclear utilities and was able to conclude that the various components in containment were functional. The licensee entered this issue into their corrective action program and plans to perform the required analysis or testing to demonstrate that all equipment that must conform to 10 CFR 50.49 is qualified for the licensee's containment spray conditions.
Corrective Action References: CR 1123701
Performance Assessment:
Performance Deficiency: The licensee's failure to qualify equipment inside containment subject to 10 CFR 50.49 to the most extreme chemical environment was a performance deficiency. Specifically, Division of Operating Reactors (DOR) Guidelines Sections 4.1.4 and 4.2.4, NUREG-0588 Section 1.3, and Institute of Electrical and Electronics Engineers (IEEE)
Std. 323-1974 Section 6.3.6 state that the equipment to be qualified should be demonstrated to be qualified (by test) for the most severe caustic environment it is expected to see given the most limiting single failure.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone. Specifically, the failure to qualify 10 CFR 50.49 equipment in containment for the most severe chemical composition of containment spray fails to ensure the reliability and capability of those components to perform their safety function when called upon during a design basis accident.
Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The finding was determined to be of very low safety significance (i.e., Green) because the finding was a deficiency affecting the qualification of a mitigating SSCs and the SSCs maintained their operability or functionality.
Cross-Cutting Aspect: Not Present Performance. No cross cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.
Enforcement:
Violation: 10 CFR 50.49(e)(3) requires, in part, that qualification of electrical equipment important to safety be based on the composition of chemicals at least as severe as that resulting from the most limiting mode of plant operation.
Contrary to this, since February 2, 1983, the licensee failed to base qualification on the composition of chemicals of electrical equipment important to safety inside containment to at least as severe as that resulting from the most limiting mode of plant operation.
Enforcement Action: This violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy.
Failure to Verify the Adequacy of Design for Aging and Accident Profile Testing Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green None (NPP) 71111.21N Systems NCV 05000338,05000339/2019010-02 Closed The inspectors identified three examples of a Green finding and associated Non-cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III when the licensee failed to verify compliance with environmental qualification (EQ) requirements in accordance with design control manual NDCM-3.3 Rev. 11, Design Verification.
Description:
Example 1:
For Units 1 and 2, the licensee failed to verify that post-accident Arrhenius calculations demonstrated the required post-accident-operating-time (PAOT) of 120 days for 300V Boston Insulated Wire (BIW) instrument cables.
Specifically, the licensees Plant Qualification Evaluation (PQE) NA-EQUAL-000-PQE-06.02-P01, Boston Insulated Wire & Cable Co. XLPE/NEO 300 V Instrument Cable. Rev. 1, verified that the site EQ requirements were met by the qualification testing documented in BIW qualification report 8714. The BIW report specified that unaged cable samples were subjected to loss-of-coolant-accident (LOCA) simulations, including PAOT, for approximately 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br />. However, the PQE evaluation of the BIW cable testing incorrectly based qualification on a test duration that lasted more than 400 days. When inspectors informed the licensee of this error, the licensee re-ran their PAOT calculation and determined that the cables could not meet their PAOT mission time.
Example 2:
For Units 1 and 2, BIW 300 V instrument cables were identified as deficient in aging for qualified life in NRC safety evaluation reports (SERs) in 1983. The licensee claimed a qualified life of 40 years, which was not supported by the cable qualification documentation.
The qualification test used unaged cable for the LOCA testing even though the cable was identified as being vulnerable to the deleterious effects of thermal aging. In response to the SERs, the licensee devised a surveillance and maintenance schedule to monitor cables and replace them as needed. Sometime after the NRC concern was closed, the licensee determined that surveillance was not practical and decided to determine a qualified life.
Since the cables were not aged prior to testing to establish a qualified life, the licensee used the vendor-supplied thermal regression line to determine a qualified life in excess of 73 years.
Fundamentally, a regression line, alone, is insufficient to determine a qualified life because it does not account for a design basis accident at the end of life plus a PAOT of 120 days.
These cables must perform their safety functions at the end of a lifetime of radiation aging, thermal aging, and other aging effects during and following design basis accidents for a minimum of 120 days. The license failed to verify that their calculational methods met their design control requirements.
Example 3:
Initially, the licensee utilized information from Conax testing (aging time, aging temperature)and a science paper published in the Society of Polymer Engineers (SPE) Journal in March 1968 (activation energy) to establish a 15.92-year qualified life (22.26 years when accurate numbers are used). Subsequently, the licensee made use of information in a Curtiss Wright System 1000 database entry (slope, intercept), which came from the same 1968 science paper, and determined that the lifetime of polysulfone was 4,103 years. As such, the licensee concluded that polysulfone is not susceptible to thermal aging. The 1968 science paper stated that downward extrapolation by 20C or 30C can be done with confidence and enable one to estimate useful life at end-use operating temperatures. The licensee failed to apply these limitations when they determined that polysulfone was not susceptible to thermal aging, indefinitely. In this manner, the licensee failed to meet the requirements of NDCM 3.3 Section 6.2.7, Design Reviews.
Because the determination that the polysulfone terminal blocks are not susceptible to thermal aging was inappropriate, the established qualified life of these components was 15.92 years (22.26 years), and because the licensee failed to realize this until it was identified by the inspectors, the components which were currently installed in the plant - which have been installed since initial operation of Unit 2 (August 1980) - are beyond their established qualified life.
Corrective Actions: The licensee entered these three issues into their corrective action program and determined that the components were functional but non-conforming.
Corrective Action References: CR 1123419, CR 1124362, CR 1124174, and CR 1124632
Performance Assessment:
Performance Deficiency: The licensee failed to verify that design inputs were selected correctly and that design inputs were correctly used in accordance with Design Control Manual NDCM 3.3 Section 6.2.7, Design Reviews.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone objective and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of the RCS. Specifically, the licensee failed to establish a qualified life and replace components (instrument cables; terminal blocks) on that interval. Keeping a component in service that has exceeded its qualified life reduces the reliability of that device to perform its safety function when called upon in a design basis accident.
Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The finding was determined to be of very low safety significance (i.e., Green) because the finding was a deficiency affecting the qualification of a mitigating SSCs and the SSCs maintained their operability or functionality.
Cross-Cutting Aspect: Not Present Performance. No cross cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.
Enforcement:
Violation: 10 CFR 50 Appendix B Criterion III requires, in part, that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.
Contrary to the above, since July 7, 1986, the licensee failed to provide design control measures for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program for their 300 V instrument cables and Connectron terminal blocks.
Enforcement Action: This violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy.
Unresolved Item Potential 50.9 Violation for Incomplete Information Submittal 71111.21 (Open) 05000338,05000339/2019010-03 N
Description:
An unresolved item (URI) was identified regarding a potential 10 CFR 50.9 violation for the licensees submittal of incomplete information. Specifically, on June 22, 2000, the licensee submitted a proposed technical specification change (Virginia Electric and Power Company North Anna Power Station Units 1 and 2 Proposed Technical Specification Changes Increased Boron Concentration, ADAMS ML003728735) intended to allow increasing boron concentration in the RWST, casing cooling tank, safety injection accumulators, and the spent fuel pool. Section 3.3.3 of the submittal was titled Post-LOCA Containment Sump and QS pH Analysis, and stated, in part: After consideration of the proposed revised boron concentrations in the RWST, CCT, and SIAs, the QS and post-LOCA sump pH analysis limits continue to be met. Specifically, the post-LOCA containment sump pH is calculated to be between 7.0 and 9.5, and the QS pH is calculated to be between 8.5 and 10.5 with at least 95% probability and confidence. Additionally, Section 3.4.2 of the submittal was titled Equipment Qualification, and stated, in part: Increasing the boron concentration to 2600 to 2800 ppm in the RWST and CCT and to 2500 to 2800 ppm in the SIAs will not adversely affect the environmental qualification of equipment in the Equipment Qualification Master List (EQML). The corrosive agent in chemical spray is primarily NaOH.
Increasing the boron concentration lowers the solution pH making it less corrosive (more neutral). Therefore, higher boron concentration limits are acceptable, even for those components qualified at a lower boron concentration. The team identified that the QS system operates in a manner which results in RWST water, with a pH as low as 4.7, spraying down containment for at least the first five minutes of a LOCA. The submittal did not include this information and was therefore incomplete. The team also identified that the licensee had not evaluated the equipment qualification for this low pH condition for components in containment Planned Closure Actions: This is an unresolved item pending NRC determination of whether the issue was material in accordance with the NRC Enforcement Policy (ADAMS ML19123A129), Section 2.3.11, Inaccurate and Incomplete Information, and the NRC Enforcement Manual (ADAMS ML18018B134), Part II, Section 1.5, Material False Statements and Completeness and Accuracy of Information.
Licensee Actions: The 50.9 aspect of this issue is an administrative issue and poses no immediate safety or security concerns. The licensee has entered the associated NCV issue into their Corrective Action Program as CR1123701 and has stated that they will address the potential 50.9 violation if the NRC determines one exists.
Corrective Action References: CR1123701
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On June 7, 2019, the inspectors presented the design basis assurance inspection (teams) inspection results to Larry Lane, Site Vice President, and other members of the licensee staff.
- On July 18, 2019, the inspectors presented the Re-Exit to Jim Jenkins, Director of Safety and Licensing at North Anna, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection Type Designation Description or Title Revision or
Procedure Date
71111.21N Calculations 11715-ES-188-1 Equipment Temperature Transient for Polysulfone Terminal 06/24/1983
Blocks
CT-REPORT- Generic Performance Requirement Assessment for Cable, Rev. 1
000-EE-0024 Terminal Block, Splice and Circuitry Performance at North
Anna and Surry Power Stations
CT-REPORT- Resolution of IEIN 82-03 Environmental Testing of Electrical Rev. 0
000-EQ-0024 Terminal Blocks for Surry and North Anna Power Stations
CT-REPORT- Usage of Electrical Terminal Blocks in Nuclear Power Rev. 0
000-EQ-0030 Stations for Surry and North Anna Power Stations
CT-REPORT- Electrical Equipment Environmental Qualification Design Rev. 0
000-EQ-0079 Basis for North Anna and Surry Power Stations
ETC-NA2004-1 Thermal Aging Analysis of Boston Insulated Wire and Cable Rev. 1
Co. Instrument Cable
ETE-NA-2017- Arrhenius Basis for Target Rock SOV Equivalent Rev. 0
0014 Temperature
NA-CALC-NFL- North Anna Quench Spray and Post-LOCA Sump pH Rev. 2
SM-0415 Calculations (Revised Safety Analysis Limits (SALs) for
RWST and CAT
NA-CALC-NFL- North Anna Quench Spray and Post-LOCA Sump pH Rev. 2
SM-0415 Calculations (Revised Safety Analysis Limits (SALs) for
RWST and CAT)
NA-CALC-NFL- North Anna Quench Spray and Post-LOCA Sump pH Rev. 2
SM-0415 (ADD- Calculations (Revised Safety Analysis Limits (SALs) for
191 pH and CAT)
NA-EQUAL-000- Environmental Zone Descriptions Rev. 27
59-EZD
NA-REPORT- Temperature Monitoring Data Rev. 1
000-EQ-0049
NE-118 Temperature Response of Environmentally Qualified Rev. 0
Equipment Due to the North Anna Power Station MSLB
Profile in Containment
SDBD-NAPS-EI System Design Basis Document for Electrical Rev. 15
Inspection Type Designation Description or Title Revision or
Procedure Date
Instrumentation and Computer Systems - North Anna Power
Station
71111.21N Calculations SM-1516 North Anna MSLB Containment Response with GOTHIC Rev. 0
System 1000 Time-Temperature Regression Calculation: Polysulfone 05/19/2019
Revision 17.0.d Tensile Impact Strength
TTR Report
71111.21N Corrective Action 1123369 Target Rock Qualification Editorial Discrepancy 05/15/2019
Documents 1123419 2019 NRC EQ DBAI - Inaccurate BIW Cable Post-Accident 05/15/2019
Resulting from Operating Time
Inspection 1123437 2019 NRC EQ DBAI - Connectron Terminal Block 05/15/2019
Discrepancies in EQ File
23503 2019 NRC EQ DBAI - Validyne Multiplexer Heat Rise 05/16/2019
23701 2019 NRC EQ DBAI - Acidic Spray Evaluation in EQ 05/21/2019
Program
23788 2019 NRC EQ DBAI - Additional Qualification References for 05/22/2019
NAMCO Limit Switches
24174 2019 NRC EQ DBAI - BIW 300V Instrumentation Cable Rev. 0
Qualification Issues
24362 2019 NRC EQ DBAI - CM-AA-EQ-10 Discrepancy for 06/06/2019
Regression Analysis
24495 2019 NRC EQ DBAI - Validyne Qualified Parts File 06/05/2019
Discrepancy
24531 2019 NRC EQ DBAI - Cable Vault Environmental Zone 06/05/2019
Inaccuracy
24534 2019 NRC EQ DBAI - Target Rock SOV Heat Rise 06/05/2019
Justification Discrepancy
24537 2019 NRC EQ DBAI - Terminal Block Discrepancies in LR- 06/05/2019
1011/LR-2011
24632 2019 NRC EQ DBAI - Issues of Concern: Connectron 06/06/2019
Terminal Blocks
24643 2019 NRC EQ DBAI - Information provided in Amendments 06/06/2019
related to Raising RWST Boron Concentration
24647 2019 NRC EQ DBAI - Terminal Blocks Installed Inside 06/06/2019
Containment Discrepancy
Inspection Type Designation Description or Title Revision or
Procedure Date
71111.21N Corrective Action 1124648 UFSAR Ch. 3 for MSLB analysis 06/06/2019
Documents
Resulting from
Inspection
71111.21N Drawings 11715-FE-7EE Multiplexer Panels Sh. 7 Technical Support Center Wiring Rev. 12
SH-001 Diagram, North Anna Power Station - Unit 1
2050-ECI- Pressurizer Spray Safety and Relief Tank Piping North Anna Rev. 6
110DA Power Station - Unit 2
13075-FE-7EC One Line Diagram Technical Support Center Multiplexer Rev. 5
System SH. 1 North Anna Power Station Unit 1
79AB-008-2 Solenoid Operated Globe Valve High Temp., High Press.
Energize to Open 1"
71111.21N Engineering NA-17-00158 Pressurizer Head Vent Equivalent Solenoid Valves/ NAPS/ Rev. 1
Changes Unit 2
CT-EQUAL-000- Test Report for Nuclear Environmental Qualification of Patel Rev. 0
GQE-34.04-G01 1/2 Inch Electrical Connector
CT-EQUAL-000- Test Report for Nuclear Environmental Qualification of Patel Rev. 0
GQE-34.04-P01 1/2 Inch Electrical Connector
EQ-0049 Evaluation of Temperature Monitoring Data from North Anna Rev. 1
Power Station
EQ-0050 Evaluation of Target Rock Report TERI 018 for the Rev. 1
Replacement Terminal Block Insulating Material
ET-CEP-02-0015 Justification and Implementation Plan for the Use of MOV Rev. 2
Long Life Grade 1 Grease North Anna and Surry Power
Stations Units 1&2
ETE-NA-2014- Evaluation of LED Replacements for Incandescent Indicator Rev. 1
0001 Lights
ETE-NA-2016- Review of EQ Documentation and Procedures for EGS P-I Rev. 1
0065 Thread Sealant for EGS Quick Disconnect Connectors
ETE-NA-2018- Acceptability of NSS3 EQ Terminal Blocks for NUREG-0588 Rev. 0
0018
NA-ENGT-000- Equipment Qualification Considerations due to Increased Rev. 0
CEP 99-0042 Boron Concentrations
71111.21N Engineering NA-EQAL-000- Plant Qualification Evaluation: Connectron NSS3 Terminal Rev. 2
Evaluations PQE-17.01-P01 Block
71111.21N Miscellaneous 45050-1 Wyle Test Report for Victoreen Two High-Range 06/12/1981
Containment Area Radiation Monitoring Systems
58666 Wyle Laboratories Nuclear Environmental Qualification of 12/16/1981
Module Case and Electric and Electronic Components of the
P.C. Boards and Plug-In Modules of the Remote Multiplexer
Units, Model Nos. MC170AD-Q2 and MC370AD-Q2 for
Validyne Engineering Corporation
9834 Validyne Qualification Specification for Remote Multiplexer Rev. 13
Units and Associated Signal Conditioner Modules
CM-AA_EQ-10 Fleet EQ Program Description Rev. 3
CO-NRC-000-00- Letter from David A. Christian to U.S. Nuclear Regulatory 06/22/2000
305 Commission, Virginia Electric and Power Company North
Anna Power Station Units 1 and 2 Proposed Technical
Specification Changes Increased Boron Concentration
CO-NRC-000-00- Letter from Leslie N. Hartz to U.S. Nuclear Regulatory 11/15/2000
556 Commission, Virginia Electric and Power Company North
Anna Power Station Units 1 and 2 Proposed Technical
Specification Changes Increased Boron Concentration
Request for Additional Information
CT-EQUAL-000- Generic Qualification Evaluation, Validyne Engineering Rev. 0
CT-EQUAL-000- Corporation, Multiplexer, Cables, Connectors, Models
GQE-06.16-G02 MC170AD-Q2 and MC370AD-Q2
CT-EQUAL-000- Plant Qualification Evaluation, Validyne Engineering Rev. 1
CT-EQUAL-000- Corporation, Multiplexer, Cables, Connectors
PQE-06.16-P02
CT-EQUAL-000- Target Rock Solenoid Valve 79AB-008BB-1 Rev. 2
PQE-35.04-P02
CT-SPEC-000- Specification for Electrical and Fiber Optic cable North Anna Rev. 7
NUC-0001 and Surry Power Stations Units 1 and 2
EGS-TR-913601- Nuclear Environmental Qualification Report of EGS 3/4 Inch Rev. B
Quick Disconnect Electrical Connector
EGS-TR-913602- Nuclear Environmental Qualification Report of EGS 1 1/2 Inch Rev. B
Quick Disconnect Electrical Connector
EQTR 3996A Qualification Test Report for the Environmental Qualification 01/19/1984
of the Target Rock Corporation Solenoid Operated Globe
Valves in Accordance with Standard Case IV Conditions
(Modified)
71111.21N Miscellaneous ET-CEP-05-0022 Review of EQ Accident Profiles in Support of GSSI-191 Rev. 0
Sump Strainers
ET-CEP-06-0016 Review of EQ Accident Profiles in Support of GSI-191 Sump Rev. 0
Strainers
ET-NAF-06-0083 NAPS MSLB/LOCA GOTHIC Analysis Data for EQ Pressure Rev. 0
and Temperature Limits
IPS-107 Conax Test Procedure and Report on Electrical Terminations 10/5/1973
Subjected to Design Basis Accident Environment
IPS-325 Conax Design Qualification Material Test Report for Rev. E
Materials Used in Conax Nuclear Products for Service in
Nuclear Power Generating Stations
NA-EQUAL-000- North Anna Environmental Zone Description Units 1 & 2 Rev. 27
59-EZD
NA-EQUAL-000- XLPE/NEO 300 V Instrument Cable Rev. 1
PQE-06.02-P01
NA-SPEC-000- Specification for 300 V Instrument Cable for North Anna Rev. 5
NAS-0128 Power Station 1975 Extension North Anna Power Station
PEI-TR-880701- EGS Test Report for Nuclear Qualification of Patel/EGS 1/2 Rev. A
connector
QDR-N-6.1 Boston Insulated Wire & Cable Co. XLPE/NEO Rev. 5
QTR 157 Generic Qualification of EA180 Series Limit Switches with Rev. 2
Receptacle and connector/Cable Assemblies
QTR 500 Qualification of NAMCO EC290-Series Receptacle Rev. 0
Assemblies and Connector/Cable Assemblies for Use in
Nuclear Power Plants in Compliance with IEEE Standards
QTR 510 Qualification of NAMCO EA180-Series Limit Switches with Rev. 0
EC290-Series Receptacle and Connector/Cable Assemblies
for Use in Nuclear Power Plants in Compliance with IEEE
Standards
QTR 82-002 Nuclear Environmental Qualification of the Remote Rev. B
Multiplexer Unit Models MC170AD-Q2 and MC370AE-Q2
and Associated PC Boards and Plug-In Modules
Safety Evaluation Safety Evaluation for Environmental Qualification of Safety- 03/24/1983
Report Related Electrical Equipment
71111.21N Miscellaneous SPE Journal Vol. Testing for Thermal Endurance: A Case History Based on March 1968
(p.52 - 55) Polysulfone Thermoplastic, T. Bugel
TR-4501 Hot Flow Continuous Energization Solenoid Valve Test 09/11/1986
Model 1032110-4
WCAP-8687, Equipment Qualification Test report Target Rock Isolation Rev. 2
Supp. 2-H10A Solenoid Valve (One Inch) (Environmental and Seismic
Testing)
71111.21N Procedures 0-ECM-0202-01 Installation and Removal of EGS Bayonet Connectors Rev. 20
0-ECM-0206-01 Installation of Lugs Rev. 13
0-ECM-2002-01 Trouble-Shooting and Repair of Namco Limit Switches Rev. 16
0-ECM-2104-01 Trouble-Shooting and Repair of Target Rock Solenoid- Rev. 17
Operated Valves
0-EPM-1503-01 Inspection of Limitorque Motor-Operated Valves Rev. 18
2-PT-210-12 Valve Inservice Inspection (RCS Gaseous Vent System) Rev. 14
CM-AA-DDC-201 Design Changes Rev. 22
CM-NA-EQ-100 Environmental Qualification Program Implementation Rev. 6
CM-NA-EQ-100 Environmental Qualification Program Implementation Rev. 6
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91-EQPM
CO-PROGRAM- Fleet EQ Program Description Rev. 3
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DNES-AA-MOV- Motor-Operated Valve Diagnostic Test Preparation and Rev. 3
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NA-EQUAL-000- Equipment Qualification Maintenance Manual Rev. 52
59-EQMM
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NA-PROCNA- Environmental Qualification Program Implementation Rev. 6
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NA-PROCNA- North Anna Breaching Environmental Zones Rev. 0
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71111.21N Procedures PDBD-NAPS Plant Design Basis Document for North Anna Power Station Rev. 6
71111.21N Self-Assessments PIR1076739 NAPS SPS EQ Program Self-Assessment 2017 2017
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