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NRC Evaluation of Changes, Tests, and Experiments Report No. 05000390/2018010 and 05000391/2018010 (Public)
	ML18124A180
Person / Time
Site:	Watts Bar
Issue date:	05/04/2018
From:	Marvin Sykes; NRC/RGN-II/DRS/EB1
To:	James Shea; Tennessee Valley Authority
References
	IR 2018010
	Download: ML18124A180 (12)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION May 4, 2018

SUBJECT:

WATTS BAR NUCLEAR PLANT - NRC EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS REPORT NUMBER 05000390/2018010 AND 05000391/2018010

Dear Mr. Shea:

On March 23, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Watts Bar Nuclear Plant Units 1 and 2, and 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.

The NRC inspectors did not identify any finding or violation of more than minor significance.

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/

Marvin D. Sykes, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos. 50-390, 50-391 License Nos. NPF-90, NPF-96

Enclosure:

Inspection Report 05000390/2018010 and 05000391/2018010,

Inspection Report

Docket Number(s): 05000390, 05000391 License Number(s): NPF-90, NPF-96 Report Number(s): 05000390/2018010, 05000391/2018010 Enterprise Identifier: I-2018-010-0053 Licensee: Tennessee Valley Authority (TVA)

Facility: Watts Bar Nuclear Plant, Units 1 and 2 Location: Spring City, TN 37381 Inspection Dates: March 19 to March 23 Inspectors: T. Fanelli, Senior Reactor Inspector (Lead),

R. Patterson, Reactor Inspector, C. Franklin, Reactor Inspector Approved By: Marvin D. Sykes, Chief Engineering Branch 1 Division of Reactor Safety Enclosure

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring TVAs performance at

Watts Bar Nuclear Plant, Units 1 and 2 by conducting a 10 CFR 50.59 inspection 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 No findings or more-than-minor violations were identified.

Additional Tracking Items Unresolved Item Potential Failure to Request NRC Approval to 71111.17T- Evaluation of (Open) Increase the OPT and OTT Response Changes, Tests, and Times Experiments URI 05000390/2018010-01 Unresolved Item Potential Failure to Perform an 50.59 71111.17T- Evaluation of (Open) Evaluation for Auxiliary Building and Changes, Tests, and Containment Isolation Experiments URI 05000390/2018010-02 Unresolved Item Potential Failure to Maintain Design 71111.17T- Evaluation of (Open) Requirements of RTDs Changes, Tests, and URI 05000390/2018010-03 Experiments

INSPECTION SCOPES

Inspections were conducted using the appropriate portions of the inspection procedure (IP) 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, performed walk downs, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.

REACTOR SAFETY

71111.17T - Evaluations of Changes, Tests and Experiments

Main article: IP 71111.17

The inspectors evaluated the following samples for compliance with Title 10, Code of Federal Regulations (CFR) 50.59, Changes, tests, and experiments, from March 19, to March 23, 2018.

10 CFR 50.59 Evaluations

(1) DCN 57326, Replace Station Air Compressors & Associated Equipment

(2) DCN 65097, Install New Backdraft Dampers for Smoke Mitigation Between the 6.9 kV B-Train Shutdown Board Room and the 480vac Shutdown Board Room

(3) DCN 66453, Unit 1 ECCS Gas Accumulation

(4) DCN 65534, Incorporate Revised Containment Integrity Analysis into Design Basis to Accommodate Dual-Unit Configuration of the GDC 5 Systems

(5) DCN 66576, Revise Design and Licensing Bases to Incorporate the Westinghouse Cobra/Trac Analysis Methodology as the Large-Break LOCA Containment Integrity Analysis of Record

(6) DCN 66577, Update MCR Dose Due to Fuel Handling Accident Caused By Increased Isolation Damper Stroke Time

(7) DCN 66327, Increased Narrow Range RTD Response Time from 8 To 9 Seconds 10 CFR 50.59 Screening/Applicability Determinations

(1) DCN 60939, Remove the ERCW Discharge Header Air Release Valves 0-ARV-67-1108, -1109, 0-TV-67-1112, -1113

(2) DCN 64044, Replace Time Delay Relays for Containment Spray Pump 1A-A AND 1B-B

(3) DCN 64501, Replace Centrifugal Charging Pump Room Cooler Coil with ASME Section III COIL

(4) DCN 65421, Valve Failed Leak Check

(5) DCN 65448, Determine the Throttle Positions for VALVES 0-FCV-067-0360 and -0362 That Limit the Flow Out of a Potential Failure of the TVA Class H ERCW Discharge Headers A and B to Acceptable Levels

(6) DCN 66557, Change Setpoint of Building Heat Expansion Tank Nitrogen Pressure Regulators

(7) DCN 66570, Create Second Architectural Opening in Unit 2 TDAFW Pump Room

(8) DCN 61035, Post-Loca Shutdown Margin Recovery

(9) DCN 65527, Replace Obsolete Valve for WBN-1-DRV-062-0521-A

(10) DCN 66387, Install Various Vent/Drain/Test Valves in System 26 and 67

(11) DCN 66508, Reduce Flow Requirements of the Boric Acid Transfer Pumps

(12) DCN 66459, Install Relays & Cable to Modify U1 Containment Purge Isolation Logic

INSPECTION RESULTS

Unresolved Item Potential Failure to Request NRC Approval 71111.17T- Evaluation (Open) to Increase the OPT and OTT Response of Changes, Tests, and Times Experiments URI 05000390/2018010-01

Description:

The reactor trips that protect from fuel damage that could result from departure from nucleate boiling around the fuel are identified as over-temperature-change-in-temperature (OTT) and over-power-change-in-temperature (OPT). The trips use the temperature from the reactor coolant systems hot legs as inputs into complex equations. In 1991, the licensee requested a license amendment to upgrade the Temperature Averaging System (TAS) and protection system to digital technology (Eagle 21 protection system). The Westinghouse topical reports (TR) for the TAS and Eagle 21 was reviewed and the TAS was approved with conditions for the RTD response times, electronic delay times, and surveillance test uncertainties in NUREG 847, the Safety Evaluation Report (SER), Supplement 8 dated January 1992.

The SER specified, that the overall response time (RTD response time plus electronics delay) for the new RdF RTDs is 0.5 second longer (6.5 vs. 6.0 seconds) than the former Rosemount RTDs. This leaves a margin of 0.5 second (7.0-6.5) between the analysis and overall RTD response time. The breakdown of components used to arrive at the overall response time is 5.5 seconds for the RTD/thermowell and a conservative electronics delay of 1.0 second. The applicant stated that it will use the loop current step response (LCSR) test to measure RTD response time. A 10-percent allowance for LCSR test uncertainty will be used to ensure an overall channel response time of 7.0 seconds or less. During initial startup testing, actions will be taken to correct any resistance temperature detector (RTD) channel with an overall response time of greater than 7.0 seconds including electronics delay and a 10-percent allowance for loop current step response test uncertainty. After any such corrective action, the channel will be retested to verify an overall response time of 7.0 seconds or less (the value assumed in pertinent safety analyses).

In 1997, licensee Design Change Notice (DCN) 39293 was implemented to increase the RTD response time. It stated, the response time requirement for OPT reactor trip was increased from 7 seconds to 8 seconds. This time includes RTDs, electronic processing, and trip circuit delays. As a result, the allowance for the sensor response time can be increased from 5.5 to 6.5 seconds. The Reactor Protection System Description, N3-99-1003, and the Technical Requirements Manual (TRM) were revised to reflect the change in response time for this channel. The change appeared to account for the 1.0 second electronic delay, but did not appear to account for the 10-percent allowance for LCSR test uncertainty, which would be derived from the RTD/thermowell delay. The uncertainty margin would appear to increase from 0.5 to 0.6 seconds. This change was implemented without NRC review and approval.

In 2015, during hot functional testing of Unit 2 TAS RTDs, the RTD/thermowell delay did not meet the 6.5s required by the TRM from the change in 1997. On May 23, 2015, DCN 66327 was implemented to increase the response time again. The DCN stated, this DCN increases the total Narrow Range RTD response time from 8 to 9 seconds while changing the sensor response time from 6.5 to 8 seconds. Westinghouse has evaluated this change in letter WBT-D-5476 and determined that existing analyses are not impacted by this change. In this new response time the 1.0 second electronic delay and 8 second RTD/thermowells delay appeared to be accounted for, but not the margin for LCRS test uncertainty. If the 10-percent allowance for LCSR test uncertainty were accounted for, the total response time would appear to increase to 9.8 seconds.

Westinghouse used a total response time of 9.0 seconds for their analyses at the direction of TVA, per WBT-TVA-3027, Revision 0, (5.10) PIN ELICB-055 Evaluation to Support a 9.0-second Total RTD Response Time, August 2015. The 10 percent LCSR uncertainty does not appear to have been included. Westinghouse letter LTR-TA-15-92, Transient Analysis Evaluation of an Increased RTD Delay Time for Watts Bar Unit 2, Rev. 0, stated, in part, due to the limiting nature of the [Steam Line Break] SLB w/ [Rod Withdrawal at Power] RWAP event, in which no margin currently exists to the departure from nucleate boiling ratio (DNBR)safety analysis limit (SAL), the inclusion of a 9.0-second total RTD response time resulted in a 0.55% DNBR penalty. For the feed water event, defined as a reduction in feedwater temperature, the Westinghouse letter stated, in part, key event results for both of the multiple-loop cases were impacted by the delay in receiving the OPT trip. While substantial margin was maintained to the DNBR limit of 1.38, the peak core heat flux values slightly exceeded the limit value of 121%. The letter concluded that the slower responding RTDs did not significantly impact the non-LOCA transient analyses and that the acceptance criteria for the events continued to be met, with the exception of the SLB w/ RWAP. However, generic DNB margin will be allocated to offset the 0.55% DNBR penalty associated with the evaluation. As such, the non-LOCA transient analyses can support operation of Watts Bar Unit 2 with a total RTD delay time of up to 9.0 seconds.

The inspectors questioned the licensee to understand why the 10-percent allowance for LCSR test uncertainty was not accounted for in the Westinghouse analyses, and to what extent it could have affected the results. In addition, the inspectors questioned whether the 10-percent uncertainty was adequate in the current installation configuration. The inspectors also questioned how the LCSR test could account for increased thermal resistance between the RTDs and the thermowells. The test may not measure the actual delay time from the hot leg across the thermowell thermal resistance to RTD. The original installations relied on specific RTD thermowell bonding to establish a predictable thermal resistance and initial response time. It is unclear how this was performed for this installation to determine the actual response time.

The 10 CFR 50.59 evaluation was performed May 22, 2016. This issue has been captured in the Corrective Action Program (CAP) as CR 1398934, Potential failure to request lic.

amendment to change OPdT/OTdT response time Unresolved Item Potential Failure to Perform a 50.59 71111.17T- Evaluation (Open) Evaluation for Auxiliary Building and of Changes, Tests, and Containment Isolation URI Experiments05000390/2018010-02

Description:

DCN 66459 added relays and wiring to change the actuation system that initiated the Auxiliary Building Isolation (ABI) and Containment Ventilation Isolation (CVI)protective functions. The new control elements (relay contacts and wiring) bypassed the Unit Solid State Protection System (SSPS) circuitry. The intent was to actuate the CVI function on an ABI actuation from the opposite unit while it was at full power operation and the other unit was in refueling mode. By bypassing the refueling units protection system and controlling the components in the refueling unit, the modification in effect made the protection systems (CVI)a shared system. The CVI was classified as part of the Engineered Safeguards Protection Systems (ESFAS). The ESFAS was not listed in the UFSAR Chapter 3.1.2 WBNP Conformance with GDCs [General Design Criteria], as shared systems under GDC 5. The UFSAR compliance with GDC 5 Sharing of Structures, Systems, and Components, specified all shared systems are sized for all credible initial combinations of normal and accident states for the two units, with appropriate isolation to prevent an accident condition in one unit from carrying into the other. The new control elements integrated in the ESFAS logic, apparently on both units. The licensee did not perform a failure modes and effects analysis to determine the negative effects that could degrade the ESFAS isolation functions when they are required to operate. The inspectors are concerned that the integration of the two ESFAS circuitry could have a detrimental effect. Additional failure modes appear to have been introduced into these systems.

The inspectors need to determine the extent to which each units protection system and CVI were exposed to additional failures including common cause failures to determine whether there could be more than a minor issue and a potential failure to perform an adequate 50.59 evaluation in accordance with NPG-SPP-09.3 Plant Mods and Engineering Change Control,Section III, was a performance deficiency. This URI, is being opened to determine whether the PD is more than minor.

This modification was complete on June 16, 2017. This issue was captured in CR 1398935, Potential violation of 10 CFR 50.59(d)(1) via DCN 66459.

Unresolved Item Potential Failure to Maintain Design 71111.17T- Evaluation (Open) Requirements of RTDs of Changes, Tests, and URI 05000390/2018010-03 Experiments

Description:

The TAS RTD receipt documents specified design criteria identified in EQOP-ESE-7 in WCAP 8587 and the EQTR WCAP-8687 Supp. 2-E07A. The installed condition of the RTDs did not appear to meet the qualification performance specifications.

Testing performed indicated that a gap existed between the RTDs and installed thermowells creating a large thermal resistance. Westinghouse document FDR-WBT-2015-01, Field Deviation Notice for Watts Bar Unit 2, indicated that the RTD/thermowell fit for the tip was outside the maximum tolerance listed in design drawing 1847E83 Rev. 4. The performance specifications were to ensure that the RTD time response across the thermowells thermal resistance was within tolerance. The higher thermal resistance increased the response time.

The deviation appeared to be accepted as is.

The Westinghouse equipment qualification data package, WCAP 8587/EQOP-ESE-7 Supplement 1 Rev. 7, page D1-5 Rev. 21, specified, in part that the performance requirements for the RTDs must meet +0.2"F repeatability; first order time response 5 seconds with well for step change of at least 20°F with a water flow of 7 ft/sec. Time response testing has been successfully performed via type testing on a sample model of this RTD. The performance requirement appears to require a 5-second RTD/thermowell response time for qualification directly measured across the thermowell medium. The inspectors are concerned that the installed RTD/thermowell configuration was not verified to meet the performance requirements for qualification.

It appeared that, due to the installation issues, the RTD response time increased, as measured by LCSR testing methods, the total delay time of 9.8 seconds. However, Westinghouse used a total response time of only 9.0 seconds for their analyses at the direction of TVA specification per WBT-TVA-3027, Revision 0, PIN ELICB-055 Evaluation to Support a 9.0-second Total RTD Response Time, August 2015. The analyses determined that some accident margins were impacted at 9.0 seconds. The inspectors questioned what the impact of the additional delay would have on the accident analyses and qualification.

The inspectors questioned how the LCSR test method could accurately account for increased thermal resistance between the RTDs and the thermowells and whether the original 10-percent uncertainty for LCSR testing was adequate in the currently installed configuration.

The validity of the LCSR method depends on how well the temperature sensor design satisfies LCSR test assumptions. The original installations relied on specific RTD thermowell bonding to establish a predictable thermal resistance and initial response time. It is unclear how the actual response time was determined this installation.

This URI is being opened to determine if a PD exists. A review of documents and specifications provided by licensee indicated that new information requests would be likely.

This issue has existed since September 2015. This issue was captured in CR 1398936, RCS Narrow Range RTDs Design and Qualification Requirements.

EXIT MEETINGS AND DEBRIEFS

The inspectors confirmed that proprietary information was controlled to protect from public disclosure. On March 23, 2018, the inspector presented the inspection results to Mr. Gordon Arent, and other members of the licensee staff.

LIST OF

DOCUMENTS REVIEWED

CORRECTIVE ACTION DOCUMENTS WRITTEN AS A RESULT OF THE INSPECTION

CR 1397855, Clean & recoat of hot tapped ERCW branch piping

CR 1398318, Metal Coffee Cup Found in 6.9 KV Board Room

CR 1398934, Potential failure to request lic. amendment to change OPT OTT response time

CR 1398935, Potential violation of 10 CFR 50.59(d)(1) via DCN 66459

CR 1398936, RCS Narrow Range RTDs Design and Qualification Requirements

PROCEDURES

1-SI-62-902-A, Boric Acid Transfer Pump 1A-A Quarterly Performance Test, Rev. 22

1-SI-62-914-A, Boric Acid Transfer Pump 1A-A Comprehensive Pump Test, Rev. 4

1-SI-62-915-A, Boric Acid Transfer Pump 1A-A Preservice Pump Test, Rev. 1

1-SI-63-10.1-A, ECCS Discharge Pipes Venting Train A Inside Containment, Rev. 23

1-SI-63-10-B, ECCS Pumps Venting-Train B, Rev. 29

1-Sl-63-10.2-A, ECCS Pumps and Discharge Pipes Venting Outside Containment, Rev. 17

99-0653, Fusible Link Test, Rev. 0

IP-ENG-001, Standard Design Process, Rev. 0

MEL datasheets associate with the SG lntellisys Controller

NPG-SPP-03.15, FSAR Management, Rev. 2

NPG-SPP-06.3, Pre-/Post-Maintenance Testing, Rev. 2

NPG-SPP-09.3, Plant Modifications and Engineering Change Control, Rev. 27

NPG-SPP-09.4, 10 CFR 50.59 Evaluations of Changes, Tests, and Experiments, Rev. 12

PR WB2-15-0005, Procurement of Backdraft Dampers, Rev. 1

Signaline Time Delay Relay Specifications for Model 330 and 360

Tl627, Failure Modes and Effects Analysis, Rev. 35

V&V documentation associated with SG lntellisys Controller

WB-DC-40-31.5, Analysis of Buried Piping, Rev. 5

WBN-SDD-N3-67-4002, ERCW Filled Piping Bases, Rev. 35

DRAWINGS

0-47W450-41, ERCW Mechanical drawing, Rev. 1

0-47W450-41, Mechanical Essential Raw Cooling Water, Rev.1

0-47w611-30-5 Units 1 & 2 Electrical Logic Diagram Ventilation System, Rev. 0

0-47w611-30-6 Units 1 & 2 Electrical Logic Diagram Ventilation System, Rev. 0

0-47w611-30-7 Units 1 & 2 Electrical Logic Diagram Ventilation System, Rev. 0

0-47W611-30-8 Units 1 & 2 Electrical Logic Diagram Ventilation System, Rev. 0

0-47W845-1, Mechanical Flow Diagram-Essential Raw Cooling Water System, Rev. 9

0-47W845-2, Mechanical Flow Diagram-Essential Raw Cooling Water System, Rev. 9

0-47W850-2, Flow Diagram Fire Protection Raw Service Water, Rev. 6

0-47W859-1, Mechanical Flow Diagram Component Cooling System, Rev. 7

0-47W859-2, Mechanical Flow Diagram Component Cooling System, Rev. 43

0-47W859-4, Mechanical Flow Diagram Component Cooling System, Rev. 1

0-47W866-3, HVAC Air Flow Diagram, Rev. 0

1-47w611-30-1 Units 1 & 2 Electrical Logic Diagram Ventilation System, Rev. 15

1-47W611-30-10 Units 1 & 2 Electrical Logic Diagram 5th Diesel Gen. Ventilation

System, Rev. 2

1-47W611-30-3 Units 1 & 2 Electrical Logic Diagram Ventilation System, Rev. 9

1-47W611-30-4 Electrical Logic Diagram Ventilation System, Rev. 25

1-47w611-88-1 Units 1 & 2 Electrical Logic Diagram Containment Isolation, Rev. 28

1-47W845-3, Mechanical Flow Diagram-Essential Raw Cooling Water, Rev. 36

1-47W845-5, ERCW System Flow Diagram, Rev. 45

1-47W846-1, Control and Service Air System Flow Diagram, Rev. 40

1-47W859-2, Mechanical Flow Diagram Component Cooling System, Rev. 41

17W302-1, ERCW Control Air & HPFP Piping, Rev. 17

17W302-6, ERCW Control Air & HPFP Piping, Rev. 22

17W302-7, ERCW Control Air & HPFP Piping, Rev. 16

2-47W611-30-1 Unit 2 Electrical Logic Diagram Ventilation System, Rev. 10

2-47W611-30-3 Unit 2 Electrical Logic Diagram Ventilation System, Rev. 7

2-47W611-30-4 Unit 2 Electrical Logic Diagram Ventilation System, Rev. 11

2-47W611-88-1 Unit 2 Electrical Logic Diagram Containment Isolation, Rev. 10

2-47W859-3, CCW System Flow Diagram, Rev. 34

2-47W859-3, Mechanical Flow Diagram Component Cooling System, Rev. 34

47W611-30-11 Units 1 & 2 Functional Control Logic Diagram Heating, Ventilation, and Air

Conditioning, Rev. 4

47W611-30-12 Units 1 & 2 Functional Control Logic Diagram Heating, Ventilation, and Air

Conditioning, Rev. 1

47W611-30-13 Units 1 & 2 Functional Control Logic Diagram Heating, Ventilation, and Air

Conditioning, Rev. 1

47W611-30-9 Units 1 & 2 Electrical Logic Diagram Ventilation System, Rev. 2

N2014-24x48-WBD-2-SUB, TVA Wing Blade Damper Submittal Drawing, Rev. 3

CALCULATIONS

B18 900511256, Demonstrated Accuracy Calculation for Main Control Room Air Intake

Radiation Monitor 0-RE-90-125, 126, and Emergency Air Intake Radiation Monitor 0-RE-90-

205, 206, Rev. 19

B18 910321251, WBN AFW System - Pump Net Positive Suction Head (NPSH) Available

Calculation

B18910123255, Analysis and Qualification of Buried ERCW Piping, Rev. 3

B18930304762, Seismic Qualification of System 030 Equipment, Rev. 11

B26 860514023, ERCW System Flow Requirements, Rev. 31

B26 890127016, ERCW System (67) Return Line Operating Mode Temperatures and

Pressures, Rev. 11

B26 890302008, Control Room Operator and Offsite Doses from a Fuel Handling

Accident, Rev. 16

B26 901220303, Design Pressures and Temperatures for the Auxiliary Feedwater

System, Rev. 13

B26 970806306, Boric Acid Concentration Analysis for BAT RWST, Rev. 11

CDQ0002902015000693, Smoke Damper Mounting Design, Rev. 0

EPMJKJ011191, AFW NPSH Analysis, Rev. 11

MDQ00106320110182, Effects of Gas Accumulation in ECCS Piping, Rev. 4

T93 090323001, Essential Raw Cooling Water (ERCW) System Pressure Drop Calculation,

Rev. 29

T93 100518008, Component Cooling System (CCS) Pressure Drop Calculation, Rev. 14

WCG1682, Analysis and Qualification of ERCW Buried Piping, Rev. 3

CORRECTIVE ACTION DOCUMENTS

663870 979323 1233104

665760 989876 1303037

665770 1007686 1307833

864803 1099357 1361911

25819 1159195

967466 1198057

WORK ORDERS

2645542 116869731 118606099

2645542 116869745 118672229

116734468 117433680 118740235

116869301 118353594 118919488

SELF-ASSESSMENT REPORTS

WBN-ENG-SSA-18-002, 50.59 Safety Assessment/Safety Evaluation Program, Rev. 0

MISCELLANEOUS DOCUMENTS

0-TI-100.011, IST Program Bases Document, Rev. 1

SDD-N3-62-4001, Chemical and Volume Control System Unit 1/ Unit 2, Rev. 38

WBN-DCD-40-36, The Classification of Piping, Pumps, Valves, and Vessels Unit 1/ Unit 2, Rev.

WCAP-17721-P, Westinghouse Containment Analysis Methodology - PWR LOCA Mass and

Energy Release Calculation Methodology, Rev. 0

WCAP 8587/ EQDP-ESE-7, Equipment Qualification Data Package Supplement 1, Rev. 7

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION II

245 PEACHTREE CENTER AVENUE NE, SUITE 1200

ATLANTA, GEORGIA 30303-1257

May 4, 2018

Mr. Joseph W. Shea

Vice President, Nuclear Regulatory Affairs

and Support Services

Tennessee Valley Authority

1101 Market Street, LP 4A

Chattanooga, TN 37402-2801

SUBJECT: WATTS BAR NUCLEAR PLANT - NRC EVALUATION OF CHANGES, TESTS,

AND EXPERIMENTS REPORT NUMBER 05000390/2018010 AND

05000391/2018010

Dear Mr. Shea:

On March 23, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection

at your Watts Bar Nuclear Plant Units 1 and 2, and 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.

The NRC inspectors did not identify any finding or violation of more than minor significance.

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/

Marvin

D. Sykes, Chief

Engineering Branch 1

Division of Reactor Safety

Docket Nos. 50-390, 50-391

License Nos. NPF-90, NPF-96

Enclosure:

Inspection Report 05000390/2018010 and 05000391/2018010,

cc: Distribution via ListServ

ML18124A180 SUNSI REVIEW COMPLETE FORM 665 ATTACHED

OFFICE RII:DRS/EB1 RII:DRS/EB1 RII:DRS/EB1 RII:DRS/EB1

SIGNATURE TNF1 RNP1 CAF4 MDS1

NAME TFANELLI RPATTERSON CFRANKLIN MSYKES

DATE 5/3/2018 5/3/2018 5/3/2018 5/ 4 /2018

E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO

U.S. NUCLEAR REGULATORY COMMISSION

Inspection Report

Docket Number(s): 05000390, 05000391

License Number(s): NPF-90, NPF-96

Report Number(s): 05000390/2018010, 05000391/2018010

Enterprise Identifier: I-2018-010-0053

Licensee: Tennessee Valley Authority (TVA)

Facility: Watts Bar Nuclear Plant, Units 1 and 2

Location: Spring City, TN 37381

Inspection Dates: March 19 to March 23

Inspectors:

T. Fanelli, Senior Reactor Inspector (Lead),

R. Patterson, Reactor Inspector,

C. Franklin, Reactor Inspector

Approved By: Marvin

D. Sykes, Chief

Engineering Branch 1

Division of Reactor Safety

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring TVAs performance at

Watts Bar Nuclear Plant, Units 1 and 2 by conducting a 10 CFR 50.59 inspection 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

No findings or more-than-minor violations were identified.

Additional Tracking Items

Unresolved Item Potential Failure to Request NRC Approval to 71111.17T- Evaluation of

(Open) Increase the OPT and OTT Response Changes, Tests, and

Times Experiments

URI 05000390/2018010-01

Unresolved Item Potential Failure to Perform an 50.59 71111.17T- Evaluation of

(Open) Evaluation for Auxiliary Building and Changes, Tests, and

Containment Isolation Experiments

URI 05000390/2018010-02

Unresolved Item Potential Failure to Maintain Design 71111.17T- Evaluation of

(Open) Requirements of RTDs Changes, Tests, and

URI 05000390/2018010-03 Experiments

INSPECTION SCOPES

Inspections were conducted using the appropriate portions of the inspection procedure (IP) 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, performed walk downs, and interviewed personnel to assess licensee

performance and compliance with Commission rules and regulations, license conditions, site

procedures, and standards.

REACTOR SAFETY

71111.17T - Evaluations of Changes, Tests and Experiments (19 Samples)

The inspectors evaluated the following samples for compliance with Title 10, Code of Federal

Regulations (CFR) 50.59, Changes, tests, and experiments, from March 19, to March 23,

2018.

CFR 50.59 Evaluations

(1) DCN 57326, Replace Station Air Compressors & Associated Equipment

(2) DCN 65097, Install New Backdraft Dampers for Smoke Mitigation Between the 6.9 kV B-

Train Shutdown Board Room and the 480vac Shutdown Board Room

(3) DCN 66453, Unit 1 ECCS Gas Accumulation

(4) DCN 65534, Incorporate Revised Containment Integrity Analysis into Design Basis to

Accommodate Dual-Unit Configuration of the GDC 5 Systems

(5) DCN 66576, Revise Design and Licensing Bases to Incorporate the Westinghouse

Cobra/Trac Analysis Methodology as the Large-Break LOCA Containment Integrity

Analysis of Record

(6) DCN 66577, Update MCR Dose Due to Fuel Handling Accident Caused By Increased

Isolation Damper Stroke Time

(7) DCN 66327, Increased Narrow Range RTD Response Time from 8 To 9 Seconds

CFR 50.59 Screening/Applicability Determinations

(1) DCN 60939, Remove the ERCW Discharge Header Air Release Valves 0-ARV-67-

1108, -1109, 0-TV-67-1112, -1113

(2) DCN 64044, Replace Time Delay Relays for Containment Spray Pump 1A-A AND

1B-B

(3) DCN 64501, Replace Centrifugal Charging Pump Room Cooler Coil with ASME

Section III COIL

(4) DCN 65421, Valve Failed Leak Check

(5) DCN 65448, Determine the Throttle Positions for VALVES 0-FCV-067-0360 and -0362

That Limit the Flow Out of a Potential Failure of the TVA Class H ERCW Discharge

Headers A and B to Acceptable Levels

(6) DCN 66557, Change Setpoint of Building Heat Expansion Tank Nitrogen Pressure

Regulators

(7) DCN 66570, Create Second Architectural Opening in Unit 2 TDAFW Pump Room

(8) DCN 61035, Post-Loca Shutdown Margin Recovery

(9) DCN 65527, Replace Obsolete Valve for WBN-1-DRV-062-0521-A

(10) DCN 66387, Install Various Vent/Drain/Test Valves in System 26 and 67

(11) DCN 66508, Reduce Flow Requirements of the Boric Acid Transfer Pumps

(12) DCN 66459, Install Relays & Cable to Modify U1 Containment Purge Isolation Logic

INSPECTION RESULTS

Unresolved Item Potential Failure to Request NRC Approval 71111.17T- Evaluation

(Open) to Increase the OPT and OTT Response of Changes, Tests, and

Times Experiments

URI 05000390/2018010-01

Description: The reactor trips that protect from fuel damage that could result from departure

from nucleate boiling around the fuel are identified as over-temperature-change-in-

temperature (OTT) and over-power-change-in-temperature (OPT). The trips use the

temperature from the reactor coolant systems hot legs as inputs into complex equations. In

1991, the licensee requested a license amendment to upgrade the Temperature Averaging

System (TAS) and protection system to digital technology (Eagle 21 protection system). The

Westinghouse topical reports (TR) for the TAS and Eagle 21 was reviewed and the TAS was

approved with conditions for the RTD response times, electronic delay times, and surveillance

test uncertainties in NUREG 847, the Safety Evaluation Report (SER), Supplement 8 dated

January 1992.

The SER specified, that the overall response time (RTD response time plus electronics

delay) for the new RdF RTDs is 0.5 second longer (6.5 vs. 6.0 seconds) than the former

Rosemount RTDs. This leaves a margin of 0.5 second (7.0-6.5) between the analysis and

overall RTD response time. The breakdown of components used to arrive at the overall

response time is 5.5 seconds for the RTD/thermowell and a conservative electronics delay of

1.0 second. The applicant stated that it will use the loop current step response (LCSR) test to

measure RTD response time. A 10-percent allowance for LCSR test uncertainty will be used

to ensure an overall channel response time of 7.0 seconds or less. During initial startup

testing, actions will be taken to correct any resistance temperature detector (RTD) channel

with an overall response time of greater than 7.0 seconds including electronics delay and a

10-percent allowance for loop current step response test uncertainty. After any such

corrective action, the channel will be retested to verify an overall response time of 7.0

seconds or less (the value assumed in pertinent safety analyses).

In 1997, licensee Design Change Notice (DCN) 39293 was implemented to increase the RTD

response time. It stated, the response time requirement for OPT reactor trip was increased

from 7 seconds to 8 seconds. This time includes RTDs, electronic processing, and trip circuit

delays. As a result, the allowance for the sensor response time can be increased from 5.5 to

6.5 seconds. The Reactor Protection System Description, N3-99-1003, and the Technical

Requirements Manual (TRM) were revised to reflect the change in response time for this

channel. The change appeared to account for the 1.0 second electronic delay, but did not

appear to account for the 10-percent allowance for LCSR test uncertainty, which would be

derived from the RTD/thermowell delay. The uncertainty margin would appear to increase

from 0.5 to 0.6 seconds. This change was implemented without NRC review and approval.

In 2015, during hot functional testing of Unit 2 TAS RTDs, the RTD/thermowell delay did not

meet the 6.5s required by the TRM from the change in 1997. On May 23, 2015, DCN 66327

was implemented to increase the response time again. The DCN stated, this DCN increases

the total Narrow Range RTD response time from 8 to 9 seconds while changing the sensor

response time from 6.5 to 8 seconds. Westinghouse has evaluated this change in letter WBT-

D-5476 and determined that existing analyses are not impacted by this change. In this new

response time the 1.0 second electronic delay and 8 second RTD/thermowells delay

appeared to be accounted for, but not the margin for LCRS test uncertainty. If the 10-percent

allowance for LCSR test uncertainty were accounted for, the total response time would

appear to increase to 9.8 seconds.

Westinghouse used a total response time of 9.0 seconds for their analyses at the direction of

TVA, per WBT-TVA-3027, Revision 0, (5.10) PIN ELICB-055 Evaluation to Support a 9.0-

second Total RTD Response Time, August 2015. The 10 percent LCSR uncertainty does

not appear to have been included. Westinghouse letter LTR-TA-15-92, Transient Analysis

Evaluation of an Increased RTD Delay Time for Watts Bar Unit 2, Rev. 0, stated, in part, due

to the limiting nature of the [Steam Line Break] SLB w/ [Rod Withdrawal at Power] RWAP

event, in which no margin currently exists to the departure from nucleate boiling ratio (DNBR)

safety analysis limit (SAL), the inclusion of a 9.0-second total RTD response time resulted in

a 0.55% DNBR penalty. For the feed water event, defined as a reduction in feedwater

temperature, the Westinghouse letter stated, in part, key event results for both of the

multiple-loop cases were impacted by the delay in receiving the OPT trip. While substantial

margin was maintained to the DNBR limit of 1.38, the peak core heat flux values slightly

exceeded the limit value of 121%. The letter concluded that the slower responding RTDs

did not significantly impact the non-LOCA transient analyses and that the acceptance criteria

for the events continued to be met, with the exception of the SLB w/ RWA

P. However,

generic DNB margin will be allocated to offset the 0.55% DNBR penalty associated with the

evaluation. As such, the non-LOCA transient analyses can support operation of Watts Bar

Unit 2 with a total RTD delay time of up to 9.0 seconds.

The inspectors questioned the licensee to understand why the 10-percent allowance for

LCSR test uncertainty was not accounted for in the Westinghouse analyses, and to what

extent it could have affected the results. In addition, the inspectors questioned whether the

10-percent uncertainty was adequate in the current installation configuration. The inspectors

also questioned how the LCSR test could account for increased thermal resistance between

the RTDs and the thermowells. The test may not measure the actual delay time from the hot

leg across the thermowell thermal resistance to RTD. The original installations relied on

specific RTD thermowell bonding to establish a predictable thermal resistance and initial

response time. It is unclear how this was performed for this installation to determine the

actual response time.

The 10 CFR 50.59 evaluation was performed May 22, 2016. This issue has been captured in

the Corrective Action Program (CAP) as CR 1398934, Potential failure to request lic.

amendment to change OPdT/OTdT response time

Unresolved Item Potential Failure to Perform a 50.59 71111.17T- Evaluation

(Open) Evaluation for Auxiliary Building and of Changes, Tests, and

Containment Isolation URI Experiments05000390/2018010-02

Description: DCN 66459 added relays and wiring to change the actuation system that

initiated the Auxiliary Building Isolation (ABI) and Containment Ventilation Isolation (CVI)

protective functions. The new control elements (relay contacts and wiring) bypassed the Unit

Solid State Protection System (SSPS) circuitry. The intent was to actuate the CVI function on

an ABI actuation from the opposite unit while it was at full power operation and the other unit

was in refueling mode. By bypassing the refueling units protection system and controlling the

components in the refueling unit, the modification in effect made the protection systems (CVI)

a shared system. The CVI was classified as part of the Engineered Safeguards Protection

Systems (ESFAS). The ESFAS was not listed in the UFSAR Chapter 3.1.2 WBNP

Conformance with GDCs [General Design Criteria], as shared systems under GDC 5. The

UFSAR compliance with GDC 5 Sharing of Structures, Systems, and Components, specified

all shared systems are sized for all credible initial combinations of normal and accident

states for the two units, with appropriate isolation to prevent an accident condition in one unit

from carrying into the other. The new control elements integrated in the ESFAS logic,

apparently on both units. The licensee did not perform a failure modes and effects analysis to

determine the negative effects that could degrade the ESFAS isolation functions when they

are required to operate. The inspectors are concerned that the integration of the two ESFAS

circuitry could have a detrimental effect. Additional failure modes appear to have been

introduced into these systems.

The inspectors need to determine the extent to which each units protection system and CVI

were exposed to additional failures including common cause failures to determine whether

there could be more than a minor issue and a potential failure to perform an adequate 50.59

evaluation in accordance with NPG-SPP-09.3 Plant Mods and Engineering Change Control,

Section III, was a performance deficiency. This URI, is being opened to determine whether

the PD is more than minor.

This modification was complete on June 16, 2017. This issue was captured in CR 1398935,

Potential violation of 10 CFR 50.59(d)(1) via DCN 66459.

Unresolved Item Potential Failure to Maintain Design 71111.17T- Evaluation

(Open) Requirements of RTDs of Changes, Tests, and

URI 05000390/2018010-03 Experiments

Description: The TAS RTD receipt documents specified design criteria identified in EQOP-

ESE-7 in WCAP 8587 and the EQTR WCAP-8687 Supp. 2-E07A. The installed condition of

the RTDs did not appear to meet the qualification performance specifications.

Testing performed indicated that a gap existed between the RTDs and installed thermowells

creating a large thermal resistance. Westinghouse document FDR-WBT-2015-01, Field

Deviation Notice for Watts Bar Unit 2, indicated that the RTD/thermowell fit for the tip was

outside the maximum tolerance listed in design drawing 1847E83 Rev. 4. The performance

specifications were to ensure that the RTD time response across the thermowells thermal

resistance was within tolerance. The higher thermal resistance increased the response time.

The deviation appeared to be accepted as is.

The Westinghouse equipment qualification data package, WCAP 8587/EQOP-ESE-7

Supplement 1 Rev. 7, page D1-5 Rev. 21, specified, in part that the performance

requirements for the RTDs must meet +0.2"F repeatability; first order time response 5

seconds with well for step change of at least 20°F with a water flow of 7 ft/sec. Time

response testing has been successfully performed via type testing on a sample model of this

RTD. The performance requirement appears to require a 5-second RTD/thermowell

response time for qualification directly measured across the thermowell medium. The

inspectors are concerned that the installed RTD/thermowell configuration was not verified to

meet the performance requirements for qualification.

It appeared that, due to the installation issues, the RTD response time increased, as

measured by LCSR testing methods, the total delay time of 9.8 seconds. However,

Westinghouse used a total response time of only 9.0 seconds for their analyses at the

direction of TVA specification per WBT-TVA-3027, Revision 0, PIN ELICB-055 Evaluation to

Support a 9.0-second Total RTD Response Time, August 2015. The analyses determined

that some accident margins were impacted at 9.0 seconds. The inspectors questioned what

the impact of the additional delay would have on the accident analyses and qualification.

The inspectors questioned how the LCSR test method could accurately account for increased

thermal resistance between the RTDs and the thermowells and whether the original 10-

percent uncertainty for LCSR testing was adequate in the currently installed configuration.

The validity of the LCSR method depends on how well the temperature sensor design

satisfies LCSR test assumptions. The original installations relied on specific RTD thermowell

bonding to establish a predictable thermal resistance and initial response time. It is unclear

how the actual response time was determined this installation.

This URI is being opened to determine if a PD exists. A review of documents and

specifications provided by licensee indicated that new information requests would be likely.

This issue has existed since September 2015. This issue was captured in CR 1398936,

RCS Narrow Range RTDs Design and Qualification Requirements.

EXIT MEETINGS AND DEBRIEFS The inspectors confirmed that proprietary information was

controlled to protect from public disclosure. On March 23, 2018, the inspector presented the

inspection results to Mr. Gordon Arent, and other members of the licensee staff.

LIST OF DOCUMENTS REVIEWED