IR 05000390/2017007
| ML17220A153 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 08/07/2017 |
| From: | Bartley J NRC/RGN-II/DRS/EB1 |
| To: | James Shea Tennessee Valley Authority |
| References | |
| IR 2017007 | |
| Download: ML17220A153 (19) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION ust 7, 2017
SUBJECT:
WATTS BAR NUCLEAR PLANT - NRC DESIGN BASES ASSURANCE INSPECTION (PROGRAMS) REPORT NUMBER 05000390/2017007 AND 05000391/2017007
Dear Mr. Shea:
On June 23, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Watts Bar Nuclear Plant, Units 1 and 2. On August 7, 2017, the NRC inspectors discussed the results of this inspection with Mr. Gordon Arent and other members of your staff. The results of this inspection are documented in the enclosed report.
NRC inspectors documented two findings of very low safety significance (Green) in this report.
These findings involved violations of NRC requirements. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the Enforcement Policy.
If you contest a violation or significance of an 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 II; the Director, Office of Enforcement; and the NRC resident inspector at the Watts Bar plant.
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 the Watts Bar plant. 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/
Jonathan H. Bartley, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos.: 50-390, 50-391 License Nos.: NPF-90, NPF-96
Enclosure:
Inspection Report 05000390/2017007 and 05000391/2017007, w/Attachment:
Supplemental Information
REGION II==
Docket Nos.: 050000390, 05000391 License Nos.: NPF-90, NPF-96 Report Nos.: 05000390/2017007, 05000391/2017007 Licensee: Tennessee Valley Authority (TVA)
Facility: Watts Bar Nuclear Plant, Units 1 and 2 Location: Spring City, TN 37381 Dates: June 6 - 23, 2017 Inspectors: E. Stamm, Senior Reactor Inspector (Lead)
T. Fanelli, Senior Reactor Inspector T. Su, Reactor Inspector Approved by: Jonathan H. Bartley, Chief Engineering Branch 1 Division of Reactor Safety Enclosure
SUMMARY
IR 05000390/2017-007, 05000391/2017-007; 06/06/2017 - 06/23/2017; Watts Bar Nuclear
Plant, Units 1 and 2; Design Bases Assurance Inspection (Programs)
The inspection activities described in this report were performed between June 6, 2017, and June 23, 2017, by three Nuclear Regulatory Commission (NRC) inspectors from Region II. Two Green NCVs were identified during this inspection. The significance of inspection findings are indicated by their color (i.e., greater than Green, or Green, White, Yellow, or Red) and determined using Inspection Manual Chapter (IMC) 0609, Significance Determination Process, (SDP) dated April 29, 2015. Cross-cutting aspects are determined using IMC 0310, Aspects Within the Cross-Cutting Areas, dated December 4, 2014. All violations of NRC requirements were dispositioned in accordance with the NRCs Enforcement Policy dated November 1, 2016.
The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision
NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green: The NRC identified a Green non-cited violation (NCV) of title 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to implement instructions to replace Namco limit switch gaskets as required to maintain environmental qualification. The licensee entered this issue into their corrective action program as CR 1309040.
The performance deficiency was determined to be more than minor because it was associated with the Equipment Performance attribute of Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, not maintaining the main steam header isolation valve limit switches in their qualified condition impacted their reliability. The inspectors determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. Because the finding was indicative of current licensee performance, the inspectors assigned the cross-cutting aspect of Documentation in the area of Human Performance [H.7] because the procedure did not contain accurate instructions related to the replacement of the gaskets. (Section 1R21.b.1)
- Green: The NRC identified a Green NCV of 10 CFR 50.49(j), for the licensees failure to maintain a complete record of qualification for Brand-Rex cables under environmental qualification binder WBNEQ-CABL-050. Specifically, the licensee could not produce a certificate of conformance related to thermal aging test data obtained from Brand-Rex.
The licensee entered this issue into their corrective action program as CR 1310230.
The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of the safety related cable systems. Specifically, the irretrievable loss of quality records that demonstrate the equipment is qualified for its application in conformance to Appendix B requirements, impacted the reliability and capability of safety-related cable systems. The inspectors determined the finding was of very low safety significance (Green) because the finding was a deficiency affecting the qualification of a mitigating SSC and the SSC maintained its operability or functionality.
This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. (Section 1R21.b.2)
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Mitigating Systems, Barrier Integrity
1R21 Design Bases Assurance Inspection (Programs)
a. Inspection Scope
The inspectors performed an inspection conducted as outlined in NRC Inspection Procedure (IP) 71111.21N, Attachment 1, Environmental Qualification (EQ) under 10 CFR 50.49 Programs, Processes, and Procedures. The inspectors assessed Watts Bars implementation of the site environmental qualification program as required by 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants. The inspectors evaluated whether Watts Bar 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 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 Watts Bars environmental qualification component documentation files; and that the components were installed in their tested configuration. Additionally, the inspectors performed in-plant walkdowns to determine whether equipment surrounding the environmental qualification component could fail in a manner that could prevent the safety function 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 are listed in the Attachment.
The inspection procedure requires the inspectors to select six to ten components to assess the adequacy of the environmental qualification program. The inspectors selected nine components for this inspection. Component samples selected for this inspection are listed below:
- WBN-1-@V-067-1876-A, valve FCV-067-0143-A (component cooling system heat exchanger A essential raw cooling water outlet flow control) supply cable
- WBN-1-FT-072-0013-G, containment spray header B flow transmitter
- WBN-1-MTR-063-0010-A, safety injection pump 1A-A motor
- WBN-1-MVOP-074-0003-A, residual heat removal pump A-A inlet flow control valve operator
- WBN-2-FSV-062-0073-A, regenerative heat exchanger letdown isolation valve B
- WBN-2-JB-293-0395-A, junction box for regenerative heat exchanger letdown isolation valves
- WBN-2-RE-090-0272-B, upper containment post-accident area radiation monitor
- WBN-2-ZS-001-0029A-A, main steam header isolation valve position switch
- WBNEQ-SPLC-002, Raychem medium voltage motor connection kits
b. Findings
.1 Failure to Replace Namco Limit Switch Gasket to Maintain EQ Qualification
Introduction:
The NRC identified a Green non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to implement instructions to replace Namco limit switch gaskets as required to maintain EQ qualification.
Description:
The main steam header isolation valve (MSIV) position limit switches, 1-ZS-001-0004A-A and 1-ZS-001-004B-A, provide open or closed signals to indication in the main control room, so that operators are aware of the valve position, and can make appropriate assessment of plant conditions. Since the MSIV limit switches need to perform their design basis function, and are located in a harsh environment, they were included within the scope of Watts Bars EQ program and were subjected to its requirements. The limit switches were Namco model EA740.
EA749-20014, Namco limit switch Maintenance and Surveillance Instructions for EA740 Series Limit Switches, Revision D, step 2.4, stated that, in order to maintain the qualification and integrity of the limit switch seals, replace the top cover gasket and screw assemblies each time the top cover is removed for any reason. Watts Bar Nuclear Plant work order (WO) 118022533 implemented the template instructions contained in the Qualification Maintenance Data Sheet (QMDS) of EQ binder 0-IZS-007 for Component ID 1-ZS-001-004A-A and 1-ZS-001-004B-A. On April 5, 2017, maintenance personnel performed steps E07 and E08 of WO 118022533. Step E07 directs the worker to verify that the top cover gasket and screw assembly are replaced if they are found to be damaged. Step E08 requires that the worker verify top cover gasket and stainless screw assembly are replaced any time the top cover is removed.
Both steps were signed off by maintenance personnel; however, the old gaskets were reinstalled. This invalidated the EQ qualification of the Namco limit switches.
The inspectors determined that the conflicting information contained in steps E07 and E08 of the QMDS was a contributing factor to the human performance event. The licensee entered this issue into their corrective action program as CR 1309040. The licensee evaluated impact of the issue and determined that the affected Namco limit switches remained operable, although they were in a degraded non-conforming status.
Analysis:
The inspectors determined that the licensees failure to replace the gasket per the instructions of WO 118022533 was a performance deficiency and a failure to meet 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. The performance deficiency was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, not maintaining the MSIV limit switches in their qualified condition impacts their reliability. The inspectors used IMC 0609, Att. 4, Initial Characterization of Findings, issued December 7, 2016, for Mitigating Systems, and IMC 0609, App. A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. Because the finding was indicative of current licensee performance, the inspectors assigned the cross-cutting aspect of Documentation in the area of Human Performance [H.7]. Cross-cutting H.7 states that the organization creates and maintains complete, accurate and up-to-date documentation. However, the procedure did not contain accurate instructions related to the replacement of the gaskets and screw assemblies, which may have contributed to the human performance event.
Enforcement:
Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, required, in part, that, activity affecting quality shall be prescribed by documented instructions, procedures, or drawings and shall be accomplished in accordance with these instructions, procedures, or drawings. WO 118022533, Step E08, required replacement of the top cover gasket anytime that the top cover is removed. Contrary to the above, on April 5, 2017, the licensee did not accomplish the replacement of the top cover gasket, an activity affecting quality, as prescribed by WO 118022533. In response to this issue, the licensee evaluated the potential condition of the limit switches and determined they remained operable. The licensee plans to correct the affected procedures and replace the top cover gaskets at the next available opportunity. This violation is being treated as an NCV consistent with section 2.3.2.a of the Enforcement Policy. The violation was entered into the licensees corrective action program as CR 1309040. (NCV 05000390/2017007-01, Failure to Replace Namco Limit Switch Gasket to Maintain EQ Qualification)
.2 Failure to Maintain an Adequate Record of Qualification
Introduction:
The NRC identified a Green NCV of 10 CFR 50.49(j), for the licensees failure to maintain a complete record of qualification for Brand-Rex cables under EQ binder WBNEQ-CABL-050. Specifically, the licensee could not produce a certificate of conformance related to thermal aging test data obtained from Brand-Rex.
Description:
The inspectors reviewed qualification binder WBNEQ-CABL-050. The inspectors noted that the licensees UFSAR committed to Regulatory Guide (RG) 1.89, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants. The RG 1.89 specified that design control measures such as testing be used to check the adequacy of design. These general requirements are contained in Criterion XI, "Test Control," and Criterion XVII, "Quality Assurance Records," of Appendix B, "Quality Assurance Criteria for Nuclear Power Plants-and Fuel Reprocessing Plants." Criterion XVII, stated, in part, that sufficient records shall be maintained to furnish evidence of activities affecting qualityand shall be identifiable and retrievable. The inspectors noted that certificates of vendor conformance to Appendix B requirements are evidence of activities affecting quality that shall be maintained.
Tab D of the binder contained two Arrhenius plots obtained from the cable vendor, Brand-Rex. This test data was used as the basis of extrapolating an activation energy for the Brand-Rex cables, and subsequently used to justify the qualified life of these safety-related, environmentally qualified cables. The first plot was the originally specified plot of record and used XLPO/XLPE insulation breakdown as the end of life failure mechanism. A second plot was later supplied by letter dated March 10, 1986, at the request of the licensee. This plot used XLPO/XLPE insulation elongation to break as the end of life failure mechanism. The inspectors questioned the quality of the data used for 136ºC on the elongation to break plot, since it was in disagreement with the insulation breakdown data, even though the two plots used same temperatures. This data also disagreed with other published elongation to break Arrhenius data for XLPE insulations at 136ºC (i.e. NUREGS, EPRI, and other test reports, etc.). The inspectors requested a certificate of conformance for these tests in order to demonstrate the objective quality of the data, however, the licensee was unable to provide any documentation. Since no certificate of conformance was available to verify the quality of the test data, the licensee documented the deficiency in CR 1310230. The inspectors determined that the lack of objective evidence for Appendix B quality called into question the reliability of the data, questioning its use in determining age or uncertainties in aging extrapolations.
Analysis:
The inspectors determined that the licensees failure to maintain a complete record of qualification for Brand-Rex cables that met the requirements of 10 CFR 50, Appendix B, as specified by RG 1.89, was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the irretrievable loss of quality records that demonstrate the equipment is qualified for its application in conformance to Appendix B requirements, affected the reliability and capability of safety-related cable systems. The inspectors used IMC 0609, Att. 4, Initial Characterization of Findings, issued October 7, 2016, for Mitigating Systems, and IMC 0609, App. A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and determined the finding was of very low safety significance (Green) because the finding was a deficiency affecting the qualification of a mitigating SSC and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance.
Enforcement:
Title 10 CFR 50.49(j), requires in part, that, a record of the qualification must be maintained in an auditable form for the entire period during which the covered item is installed in the nuclear plant to permit verification that each item of electric equipment important to safety is qualified for its application. Contrary to the above, since July 28, 1995, the licensee failed to maintain a complete record of qualification in WBNEQ-CABL-050 for Brand-Rex cables. Specifically, the licensee could not produce a certificate of conformance to demonstrate objective evidence of the quality for thermal aging test data obtained from Brand-Rex. This violation is being treated as an NCV consistent with section 2.3.2.a of the Enforcement Policy. The violation was entered into the licensees corrective action program as CR 1310230. (NCV 05000390/2017007-02 and 05000391/2017007-02, Failure to Maintain an Adequate Record of Qualification)
.3 Unresolved Item (URI) 05000390, 391/2017007-03, Potential Failure to Address
Environmental Qualification of Brand-Rex Cables
Introduction:
The NRC opened a URI to review the adequacy of the licensees justification for the qualified life of Brand-Rex cables, documented in EQ binder WBNEQ-CABL-050.
Description:
The inspectors reviewed qualification binder WBNEQ-CABL-050 for Brand-Rex cables. Tab C of the binder contained WBNAPS2-089, EQ Calculation for Brand-Rex Cable Insulation, which determined the qualified life and evaluated the operability during design basis accidents for materials used in Brand-Rex cable insulation for plant service environments. Tab D of the binder documented three cable qualification test reports, each of these specified that the qualification standard of record was IEEE 323-1974, IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations, in accordance with the Updated Final Safety Analysis Report (UFSAR). Tab D also contained two Arrhenius plots obtained from the cable vendor. The first plot was the originally specified plot of record and used XLPO/XLPE insulation breakdown as the end of life failure mechanism. A second plot was later supplied by letter dated March 10, 1986, at the request of the licensee. This plot used XLPO/XLPE insulation elongation to break as the end of life failure mechanism. Both plots used temperatures at 113ºC, 121ºC, 136ºC, and 150ºC, and indicated that a single test sample was evaluated at each temperature periodically until the end of life was identified. From these documents, the inspectors determined that the licensee qualified the cables by a combination of type test and analysis supported by type test data in accordance with IEEE 323-1974.
The inspectors reviewed the data from these plots and determined they had activation energies of 1.37eV for the insulation breakdown plot and 1.48eV for the elongation to break plot. The qualification binder did not use data from the first plot, and eliminated the 150ºC test data point from the second plot. The result was an activation energy of 2.13eV, which indicated a 40-year life for the insulation. The criteria for analysis in IEEE 323-1974, Section 6.5, specified, in part, that for extrapolation, the modes of failure produced under intensified or accelerated environmental or other influences shall be the same as those predicted under the required service conditions. The inspectors believe that insulation breakdown was a known predicted failure mechanism. Section 6.5 also specified, in part, that a valid mathematical model is required and it shall be based upon established principles and verifiable test data. The inspectors determined that the licensee extrapolated an invalid mathematical model by using an invalid treatment of the test data. In the licensees development of the mathematical model for this XLPE/XLPO insulation, they removed test data that closely agreed with published 150ºC test data, and they used test data that differed from published 136ºC test data. When asked, the licensee could not adequately justify, verify, or validate that their treatment of the test data met 10 CFR 50.49 requirements. The inspectors determined that this treatment altered the extrapolation of the mathematical model and produced a 2.13eV activation energy that differed from published activation energies for this type of insulation. Because of this treatment and the lack of verification, validation, and justification, the inspectors believe that the licensees mathematical model did not meet the specifications for analysis in IEE 323-1974.
The standard IEEE 323-1974, Section 6.5.4, specified that, the qualified life shall be based upon the known limits of extrapolation of the time dependent environmental effects if an accelerated aging test was used to determine the mathematical model, which required that the certainty of the statistical analysis bound the qualified life of the components for qualification.
The limits of extrapolation was described as confidence limits in IEEE 101-1972, IEEE Guide for the Statistical Analysis of Thermal Life Test Data, which is referenced in IEEE 323-1974, Section 6.3.3, as one of the established principles at the time of these Arrhenius extrapolations. The IEEE 101-1972 standard is also referenced in NUREG 0588, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment, (ADAMS Accession No. ML031480402) as an established principle in the resolutions to questions about aging in Section 4. The inspectors believe that the licensee did not account for uncertainties, which must be quantified in their extrapolations in accordance with established principles and NUREG 0588. In addition, established principles in consensus standards specified that the certainty of the extrapolations are strongly dependent on the data for the various test temperatures, spaced at a minimum of 10ºC apart, and the total number of samples for each test temperature. For the uncertainties, either fewer test temperatures or fewer samples could produce larger uncertainties, and thus wider confidence limits that could reduce the qualified life. The inspectors noted that the Arrhenius plot used by the licensee at 113ºC and 121ºC were separated by only 8ºC, which was less than required minimum for accurate extrapolations. The inspectors noted that the licensees extrapolations could contain wide confidence limits based on the limited test temperatures and samples.
In response to these concerns, the licensee provided NUREG-0847, Supplement No. 15, Safety Evaluation Report related to the operation of Watts Bar Nuclear Plant Units 1 and 2, Appendix DD, page 36, dated June 1995, which documented an audit of the licensees cable binder WBNEQ-CABL-050 by the NRC and their contracted audit team, Idaho National Engineering Laboratory (INEL), which concluded in March 1995. The inspectors noted that this binder had been periodically updated since the audit, including the addition of calculation, WBNAPS2-089, EQ Calculation for Brand-Rex Cable Insulation. Further review of the EQ binder reviewed by the NRC/INEL audit is needed to determine whether the conclusions were based on information similar to that which is included in the current binder. This URI is opened to determine if a performance deficiency exists and it is identified as URI 05000390/2017007-03 and 05000391/2017007-03, Potential Failure to Address Environmental Qualification of Brand-Rex Cables.
.4 URI 05000390, 391/2017007-04, Potential Failure to Justify Qualification of O-Rings by
Introduction:
The inspectors opened a URI to review the adequacy of critical characteristics identified for commercial grade dedication of O-rings used in an EQ application.
Description:
In 2008, TVA began to dedicate commercial grade O-rings from Parker Hannifin through a vendor, Qualtech. The O-rings are used in components that must operate in harsh environments and thus are subject to the strict requirements in 10 CFR 50.49. The dedication process used Fourier Transfer Infrared Spectroscopy (FTIR) to identify chemical characteristics, durometer to measure rubber hardness, and visual inspection to verify physical dimensions and color. When asked if this FTIR or durometer test can verify activation energy or other aging mechanisms, the licensee stated, that FTIR cannot by itself validate activation energy or thermal aging mechanisms. Further review is needed to determine the limitations involved with the accuracy of FTIR and durometer testing. The inspectors could not verify that this dedication process specified the required critical characteristics to ensure compliance with 10 CFR 50.49. The inspectors further questioned how uncertainties could be determined and evaluated from these techniques. Since the licensee is relying on this dedication process, the components were not aged and tested in accordance with 10 CFR 50.49 requirements. The licensee contends that the verification of the material by FTIR and durometer analysis provided reasonable assurance that the original qualification test reports by Westinghouse remained applicable to the dedicated O-rings.
This URI is opened to determine if a performance deficiency exists and it is identified as URI 05000390/2017007-04 and 05000391/2017007-04, Potential Failure to Justify Qualification of O-Rings by Commercial Grade Dedication.
.5 URI 05000390, 391/2017007-05, Potential Failure to Address Environmental
Qualification of Barton Transmitters
Introduction:
The inspectors opened a URI to review the adequacy of the licensees justification for the qualified life of Barton Model 764 differential pressure transmitters, documented in EQ binder WBNEQ-XMTR-001.
Description:
The inspectors reviewed EQ binder WBNEQ-XMTR-001 for Barton 764 differential pressure transmitters. The qualification standards of record were IEEE 323-1974 and NUREG 0588, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment, Revision 1. The pressure transmitters contain a cover and a housing O-ring that are specified to be replaced following calibration at least every two years to maintain qualification, and internal mineral oil seal O-rings for the sensor diaphragms that are not regularly replaced. The binder specified that the transmitter internal electronics and the O-rings were qualified for 40 years. The transmitters in question are in lower containment specified at 120°F and exposed to the maximum accumulated ambient radiation absorbed dose of 2E7 Rad over 40 years.
Westinghouse (WEC) aged the transmitters for a 0.5eV activation energy at 125°C including the electronics inside and the O-rings. The 0.5eV activation energy was specified for conservatism to account for the known synergistic effects of simultaneous thermal exposure and radiation exposure. This synergism can be significant and was also identified as significant for the O-ring material in EQ NUREGs (e.g. NUREG/CR-3629, NUREG/CR-3588) and various research papers.
During the 0.5eV EQ testing, WEC noted that the transmitter O-rings (all) failed at 1,673 hours0.00779 days <br />0.187 hours <br />0.00111 weeks <br />2.560765e-4 months <br />, which equated to less than six years of aging at 120°F. Westinghouse chose a new activation energy (0.92eV) for the O-rings, but continued to use 0.5eV for the electronics. The transmitter oil seal O-rings were replaced with new ones. Then, the transmitters with the original electronics and new oil seal O-rings were aged for an additional 350 hours0.00405 days <br />0.0972 hours <br />5.787037e-4 weeks <br />1.33175e-4 months <br />, while the cover and housing O-rings were only replaced with new ones during the final 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />. The inspectors evaluated the end of the accelerated aging processes for 120°F. The inspectors noted that the electronics aging was accelerated to less than 7.5 years total at 0.5ev. The oil seal O-ring aging was accelerated to 22.6 years at 0.92eV. The cover and housing O-ring aging was accelerated to 11 months at 0.92eV.
Further information from the licensee is needed to evaluate the justification for the 40-year qualified life of the transmitters. Specifically, the justification for the deviation from the original 0.5eV activation energy provided by WEC and the demonstration that synergistic effects and uncertainties were accounted for when deviating from this activation energy. This URI is opened to determine if a performance deficiency exists and it is identified as URI 05000390/2017007-05 and 05000391/2017007-05, Potential Failure to Address Environmental Qualification of Barton Transmitters.
OTHER ACTIVITIES
4OA6 Meetings, Including Exit
On June 23, 2017, the inspectors presented the inspection results to Mr. Jesse James and other members of the licensees staff. On August 7, 2017, a re-exit meeting was conducted via teleconference to present the final inspection results to Mr. Gordon Arent and other members of the licensees staff. Proprietary information that was reviewed during the inspection was returned to the licensee or destroyed in accordance with prescribed controls.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
- D. Acselrod, Lead, Mechanical Design
- G. Arent, Plant Support Director
- D. Baker, Nuclear Engineer, Licensing
- E. Bradley, I&C / Digitial Equipment Corporate Program Manager
- M. Casner, Site Recovery Team Manager
- S. Connors, Plant Manager
- K. Decker, Nuclear Engineer, Licensing
- B. Jenkins, Director, WBN Engineering
- D. Lee, Sr. Manager, Design Engineering
B. Merriman BOP Systems Engineer
- J. Polickoski, Licensing Corporate Functional Area Manager
- T. Ritchie, Electrical Foreman
NRC personnel
- J. Nadel, Senior Resident Inspector, Watts Bar
- J. Hamman, Resident Inspector, Watts Bar
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened &
Closed
- 05000390/2017007-01 NCV Failure to Replace Namco Limit Switch Gasket to Maintain EQ Qualification (Section 1R21.b.1)
- 05000390, 391/2017007-02 NCV Failure to Maintain an Adequate Record of Qualification (Section 1R21.b.2)
Opened
- 05000390, 391/2017007-03 URI Potential Failure to Address Environmental Qualification of Brand-Rex Cables (Section 1R21.b.3)
- 05000390, 391/2017007-04 URI Potential Failure to Justify Qualification of O-Rings by Commercial Grade Dedication (Section 1R21.b.4)
- 05000390, 391/2017007-05 URI Potential Failure to Address Environmental Qualification of Barton Transmitters (Section 1R21.b.5)