IR 05000498/2015007
| ML15205A370 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 07/24/2015 |
| From: | Thomas Farnholtz Region 4 Engineering Branch 1 |
| To: | Koehl D South Texas |
| G. George | |
| References | |
| IR 2015007 | |
| Download: ML15205A370 (23) | |
Text
UNITED STATES uly 24, 2015
SUBJECT:
SOUTH TEXAS PROJECT UNITS 1 AND 2 - NRC EVALUATIONS OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000498/2015007 and 05000499/2015007
Dear Mr. Koehl:
On June 18, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the South Texas Project, Units 1 and 2. On June 18, 2015, the NRC inspectors discussed the initial results of this inspection with Mr. Gerald Timothy Powell, Site Vice President, and other members of your staff. On July 7, 2015, the final results of this inspection were discussed with Mr. Jim Connolly, General Manager, Engineering, and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report.
The NRC inspectors did not identify any findings or violations during this inspection.
In accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding, of the NRCs Rules of Practice, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS) component of the NRCs Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Thomas R. Farnholtz, Chief Engineering Branch 1 Division of Reactor Safety Dockets: 50-498; 50-499 Licenses: NPF-76; NPF-80 Enclosure: Inspection Report 05000498/2015007 and 05000499/2015007 w/Attachment: Supplemental Information
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket(s): 50-498; 50-499 License(s): NPF-76; NPF-80 Report(s): 05000498/2015007; 05000499/2015007 Licensee: STP Nuclear Operating Company Facility: South Texas Project, Units 1 and 2 Location: Wadsworth, TX Dates: June 1, 2015 to July 7, 2015 Inspectors: G. George, Senior Reactor Inspector, Engineering Branch 1, Region IV, DRS W. Sifre, Senior Reactor Inspector, Engineering Branch 1, Region IV, DRS J. Braisted, Reactor Inspector, Engineering Branch 1, Region IV, DRS I. Khan, Reactor Inspector, Engineering Branch 3, Region III, DRS Approved By: T. Farnholtz, Chief, Engineering Branch 1 Division of Reactor Safety, Region IV-1- Enclosure
SUMMARY
IR 05000498/2015007; 05000499/2015007; 06/01/2015 - 06/19/2015; South Texas Project
Units 1 and 2; Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications.
This report covers a two-week announced baseline inspection on evaluations of changes, tests, and experiments and permanent plant modifications. The inspection was conducted by Region IV based engineering inspectors. No findings of more-than-minor significance were identified.
The significance of most findings is indicated by their color (i.e., greater than Green, or Green,
White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Cross-cutting aspects were determined using IMC 0310, Aspects Within the Cross-Cutting Areas. Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy, dated February 4, 2015. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 5, dated February 2014.
NRC-Identified Findings
and Self-Revealed Findings No findings were identified.
Licensee-Identified Violations
No findings were identified.
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and Emergency Preparedness
1R17 Evaluations of Changes, Tests, and Experiments and Permanent Plant
Modifications (71111.17T)
.1 Evaluations of Changes, Tests, and Experiments
a. Inspection Scope
The inspectors reviewed 7 evaluations performed pursuant to Title 10, Code of Federal Regulations (CFR), Part 50, Section 59, to determine whether the evaluations were adequate and that prior NRC approval was obtained as appropriate. The inspectors also reviewed 25 screenings, where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. The inspectors reviewed these documents to determine if:
- the changes, tests, and experiments performed were evaluated in accordance with 10 CFR 50.59 and that sufficient documentation existed to confirm that a license amendment was not required;
- the safety issue requiring the change, tests and experiment was resolved;
- the licensee conclusions for evaluations of changes, tests, and experiments were correct and consistent with 10 CFR 50.59; and
- the design and licensing basis documentation was updated to reflect the change.
The inspectors used, in part, Nuclear Energy Institute (NEI) 96-07, Guidelines for 10 CFR 50.59 Implementation, Revision 1, to determine acceptability of the completed evaluations and screenings. The NEI document was endorsed by the NRC in Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, dated November 2000. The list of evaluations, screenings and/or applicability determinations reviewed by the inspectors is included as an to this report.
This inspection constituted 7 samples of evaluations and 25 samples of screenings and/or applicability determinations as defined in IP 71111.17-04.
b. Findings
No findings were identified.
.2 Permanent Plant Modifications
a. Inspection Scope
The inspectors reviewed 11 permanent plant modifications that had been installed in the plant during the last three years. The modifications were selected based upon risk significance, safety significance, and complexity. The inspectors reviewed the modifications selected to determine if:
- the supporting design and licensing basis documentation was updated;
- the changes were in accordance with the specified design requirements;
- the procedures and training plans affected by the modification have been adequately updated;
- the test documentation as required by the applicable test programs has been updated; and
- post-modification testing adequately verified system operability and/or functionality.
The inspectors also used applicable industry standards to evaluate acceptability of the modifications. The list of modifications and other documents reviewed by the inspectors is included as an Attachment to this report.
This inspection constituted 11 permanent plant modification samples as defined in IP 71111.17 04.
.2.1 Replacement of Unit 1 and 2 Reactor Water Storage Tank Level Transmitters
The inspectors reviewed design change package 11-194-27, implemented to replace the refueling water storage tank level transmitter B2SILT0932. The refueling water storage tank level transmitters design function is to provide post-accident monitoring of tank level, setpoint alarms for the control room, and an interlock for emergency core cooling recirculation mode. The previous Barton level transmitter was subject to abnormal drift.
This design change package involved replacing the previous level transmitter with an environmentally and seismically qualified Rosemount transmitter. Finally, the change included an evaluation of the instrument uncertainty of the Rosemount transmitters effect on the level setpoints of the system. The inspectors did not identify any concerns with the design change package.
.2.2 Unit 2, Train A, Safety Injection Test Header Air Operated Valve Replacement
The inspectors reviewed design change package 03-14479, implemented to replace the train A safety injection header air operated valve. The existing valve was prone to leaking resulting in significant maintenance and operational burden. The valve was required to be leak tight to maintain the train A safety injection accumulators operable.
The licensee determined that repair of the valve was not a viable option because the valve was no longer manufactured. The engineering change involved replacing the air operated valve with a simpler, more reliable, manually operated valve. The change included an evaluation for the effects of the change in actuator on the accident analyses as well as normal operations. The licensee determined that there was no effect on the accident analyses and that the new more leak-tight valve resulted in a reduction in operator burden. The inspectors did not identify any concerns with the design change package.
.2.3 Replace Unit 1 and 2 Train B Emergency Safety Feature Transformers with New Load
Tap Changing Transformers and Voltage Regulating Controllers The inspectors reviewed design change package 04-11502, implemented to replace the Unit 1 and 2 Train B emergency safety feature transformers with load tap changing transformers with voltage regulating controllers. The load tap changers were introduced to ensure that the safety bus voltages would be maintained within their required band regardless of variations in switchyard voltages. This modification was implemented to address potential tripping of the primary source by the degraded voltage relays. The inspectors reviewed the analyses associated with the modification. The licensee determined that failure modes associated with the load tap changer were bounded by the failure modes associated with the original transformers with analog controllers. The evaluation also verified that no new operator actions were required. The inspectors did not identify any concerns with the design change package.
.2.4 Steam Generator Power Operated Relief Valves (PORVs) Fail Closed Modification
The inspectors reviewed design change package 08-9595, implemented to ensure that the steam generator PORVs close on a loss of power as described in the final safety analysis report (as updated). Upon review, in response to Regulatory Issue Summary 2005-29, Anticipated Transients That Could Develop into More Serious Events, the licensee determined that the steam generator PORVs would continue to operate following a loss of Class 1E power event until the operator hydraulic pressure decreased to 1500 psi. At this pressure, the PORV would fail as-is. In order to address this issue, the licensee modified the PORV control circuit to ensure that the PORV would close on a loss of Class 1E power. This modification required the reconfiguration of a control solenoid such that the PORV would automatically close on a loss of power as described in the final safety analysis report (as updated). The inspectors reviewed the analyses and post-modification verification and testing associated with this change. The inspectors did not identify any concerns associated with the design change package.
.2.5 Replacement of Residual Heat Removal Pump 2B Motor
The inspectors reviewed design change package 11-8615, implemented to replace the residual heat removal pump 2B motor. The licensee identified a reduction in insulation resistance in the installed residual heat removal pump 2B motor and elected to replace the motor in refueling outage 2RE15. The replacement motor was made by a different vendor and had slightly different full load current, locked rotor current, and locked rotor torque. The replacement motor also had different physical characteristics, smaller and lighter. The licensee determined that the replacement motor specification were bounded by the existing analyses. The inspectors reviewed the motor replacement analysis. The inspectors did not identify any concerns associated with the design change package.
.2.6 Update of Calculations Affecting Auxiliary Feedwater Storage Tank Volume
The inspectors reviewed design change package 10-23767-5, implemented to update design basis information based on changes made to multiple calculations affecting the auxiliary feedwater storage tank volume. The licensee identified that a calculation had incorrectly incorporated volume requirements for the steam generators and had failed to include a portion of the volume required from the decay heat load. The auxiliary feedwater storage tank supplies water to the auxiliary feedwater pumps and must contain a minimum volume of water equivalent to maintain the plant in hot standby for at least four hours and cooldown the primary system to the residual heat removal cut-in temperature with or without onsite power and an assumed single-failure. The inspectors reviewed the affected calculations, which included the long term cooling analysis, auxiliary feedwater storage tank volume and setpoints, and the auxiliary feedwater storage tank level instrument loop uncertainty calculation. The affected calculations did not result in changes to the related technical specification and emergency operating procedure setpoints. Additionally, the inspectors interviewed the engineers responsible for the change. The inspectors did not identify any concerns with the design change package.
.2.7 Replacement of Auxiliary Feedwater Turbine Trip and Throttle Valve Body-to-Bonnet
Gasket The inspectors reviewed design change package 13-13224-31, implemented to replace the auxiliary feedwater turbine trip and throttle valve body-to-bonnet gasket. The licensee identified that the original gasket was not rated for the temperatures experienced during turbine operation. The design change package DCP provided the technical justification, critical design attributes, design inputs, and margin impacts for use of the new gasket, which the inspectors reviewed. Additionally, the inspectors reviewed vendor documentation and the implementing work order and interviewed the engineers responsible for the change. The inspectors did not identify any concerns with the design change package.
.2.8 Update of Calculations Affecting Spent Fuel Pool Level Indication Instrumentation
The inspectors reviewed a supplement to design change package 12-12320-10, implemented to install reliable spent fuel pool level indication that can be used in responding to beyond design basis external events. Supplement 1 to design change package 12-12320-10, specifically involved the revision of calculations to add conduit attachment loads as a result of rerouting the conduits, the revision of another calculation to update the methodology used to calculate the effect of the horn cover for the sloshing loads on the horn support, and the installation of a resistor to convert the analog input from the spent fuel pool level indication instrumentation to a voltage signal for the plant computer. The inspectors reviewed the affected calculations and installation work orders. Additionally, the inspectors walked-down the spent fuel pool level indication system and interviewed the engineers responsible for the change. The inspectors did not identify any concerns with this supplement to the design change package.
.2.9 Adjustment of High Range Radiation Monitor Loss-of-Counts Time Delay
The inspectors reviewed a supplement to design change package 04-8245-33, implemented to replace the original high range radiation monitor cables with cables less susceptible to the effects of thermally induced current. Supplement 2 to design change package 04-8245-33, involved the adjustment of the high range radiation monitors loss-of-counts time delay, which was necessary due to how the new cables affected the response of the high range radiation monitors during certain plant conditions.
Specifically, the change eliminated a false indication of a common-mode failure of the high range radiation monitors during a high energy line break. Additionally, the inspectors reviewed the changes 10 CFR 50.59 evaluation, emergency operating procedures and basis documents, reactor containment building thermally induced current analysis, operator training on thermally induced current effects, and interviewed the engineers responsible for the change. The inspectors did not identify any concerns with this supplement to the design change package.
.2.10 Replace Safety Related Inverters EIV001 and 002
The inspectors reviewed design change package 05-10905-4, implemented to replace 25 kVA safety-related inverters/rectifiers. The inverters/rectifiers provide 120 Vac power for vital instrumentation and control loads. The existing inverters were obsolete and replaced with replacement Class 1E qualified inverters. This design change involved replacement of the existing inverter/rectifier with a Class 1E replacement and installation of a static transfer switch. The inspectors did not identify any concerns with the design change package.
.2.11 Eliminate Relay and Contact Race Issue for D1AFMOV0143
The inspectors reviewed design change package 14-6074-1, implemented to eliminate a timing issue associated with the operation of the auxiliary feedwater steam admission valve and auxiliary feedwater steam admission bypass valve. The contacts used for control of the two valves were located on different limit switch rotors, which resulted in a timing issue caused by the rotors and contacts operating at slightly different times. This design change moved the contacts used for control of the two valves to the same rotor to eliminate the timing issue. The inspectors did not identify any concerns with the design change package.
b. Findings
No findings were identified.
OTHER ACTIVITIES
4OA2 Problem Identification and Resolution
.1 Review of Corrective Action Program Documents
a. Inspection Scope
The inspectors reviewed corrective action program documents that identified or were related to 10 CFR 50.59 program and permanent plant modifications. The inspectors reviewed these documents to evaluate the effectiveness of corrective actions related to permanent plant modifications and evaluations of changes, tests, and experiments. In addition, corrective action documents written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problems into the corrective action system. The list of specific corrective action documents that were sampled and reviewed by the inspectors are listed in the to this report.
b. Findings
No findings were identified.
4OA6 Meetings
.1 Exit Meeting Summary
On June 18, 2015, the inspectors presented the initial inspection results to Mr. Gerald Timothy Powell, Site Vice President, and other members of the licensees staff. On July 7, 2015, the inspectors presented the final inspection results to Mr. Jim Connolly, General Manager, Engineering, and other members of the licensees staff. The licensee acknowledged the results as presented. While some proprietary information was reviewed during this inspection, no proprietary information was included in this report.
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
- C. Pham, Quality Assurance Assessor
- D. Chamberlain, Supervisor, Civil Design Engineering
- D. Gore, Supervisor, Reactor Analysis
- D. Rencurrel, Senior Vice President Operations
- G. E. Shinzel, Supervising Engineer
- G.T. Powell, Site Vice President
- H. Le, Engineer Licensing Consultant
- J. Connolly, General Manager, Engineering
- J. Cook, Coordinator, Design Engineering
- L. Sterling, Supervisor, Licensing
- M. Berg, Manager, Design Engineering, Test, and Programs
- P. Travis, Environmental Supervisor
- Q. Huynh, Mechanical Engineer, Design Engineering
- R. D. Savage, Engineer Specialist Licensing Consultant
- R. Dunn, Manager, Nuclear Fuel & Analysis
- R. Engen, Manager, Engineering Projects
- R. Gonzalez, Engineer Senior
- R. Kersey, Supervisor, Design Engineering
- R. Lane, Operations
- T. Jacobs, Design Engineering
NRC Personnel
- A. Sanchez, Senior Resident Inspector
- N. Hernandez, Resident Inspector
- S. Janicki, Project Engineer
- M. Watford, General Engineer, Plant Licensing Branch - 4, NRR
Attachment 1