IR 05000390/2012008: Difference between revisions

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{{Adams
{{Adams
| number = ML12165A186
| number = ML112790114
| issue date = 06/13/2012
| issue date = 10/06/2011
| title = IR 05000390-12-008, 01/30/2012 - 05/15/2012, Watts Bar Nuclear Plant, Unit 1, Component Design Bases Inspection
| title = Notification of Watts Bar Nuclear Plant, Unit 1, Component Design Bases Inspection - NRC Inspection Report 05000390-12-008
| author name = Nease R
| author name = Desai B
| author affiliation = NRC/RGN-II/DRS/EB1
| author affiliation = NRC/RGN-II/DRS/EB1
| addressee name = Shea J
| addressee name = Krich R
| addressee affiliation = Tennessee Valley Authority
| addressee affiliation = Tennessee Valley Authority
| docket = 05000390
| docket = 05000390
| license number = NPF-090
| license number = NPF-090
| contact person = Linda K. Gruhler 404-997-4633
| contact person =  
| document report number = IR-12-008
| document report number = IR-12-008
| document type = Inspection Report, Letter
| document type = Letter
| page count = 42
| page count = 7
}}
}}


Line 18: Line 18:


=Text=
=Text=
{{#Wiki_filter:June 13, 2012
{{#Wiki_filter:October 06, 2011


==SUBJECT:==
==SUBJECT:==
WATTS BAR NUCLEAR PLANT - NRC COMPONENT DESIGN BASES INSPECTION - INSPECTION REPORT 05000390/2012008
NOTIFICATION OF WATTS BAR NUCLEAR PLANT, UNIT 1, COMPONENT DESIGN BASES INSPECTION - NRC INSPECTION REPORT 05000390/2012008


==Dear Mr. Shea:==
==Dear Mr. Krich:==
On, May 15, 2012, U. S. Nuclear Regulatory Commission (NRC) completed an inspection at your Watts Bar Nuclear Plant, Unit 1. The enclosed inspection report documents the inspection results, which were discussed on May 15, 2012, with Mr. David Gronek and other members of your staff.
The purpose of this letter is to notify you that the U.S. Nuclear Regulatory Commission (NRC)
Region II staff will conduct a component design bases inspection at your Watts Bar Nuclear Plant during the weeks of January 30 - February 3, February 13 - 17, and February 27 - March 2, 2012. The inspection team will be led by Shane Sandal, a Senior Reactor Inspector from the NRC's Region II Office. This inspection will be conducted in accordance with the baseline inspection procedure, Procedure 71111.21, Component Design Bases Inspection, issued December 6, 2010.


The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your licenses. The team reviewed selected procedures and records, observed activities, and interviewed personnel.
The inspection will evaluate the capability of risk significant/low margin components to function as designed and to support proper system operation. The inspection will also include a review of selected operator actions, operating experience, and modifications.


This report documents three NRC identified findings of very low safety significance (Green),
From an email reply on September 29, 2011, Mr. Sandal confirmed with Mr. Riedl of your staff, arrangements for an information gathering site visit and the three-week onsite inspection. The schedule is as follows:
which were determined to involve violations of NRC requirements. The NRC is treating these violations as non-cited violations consistent the NRC Enforcement Policy. If you contest these non-cited violations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-001; with copies to the Regional Administrator Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at Watts Bar. Further, if you disagree with the cross-cutting aspect assigned to any finding 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 Regional Administrator, Region II, and the NRC Resident Inspector at Watts Bar. The information you provide will be considered in accordance with Inspection Manual Chapter 0305.
* Information gathering visit: Week of January 9 - 13, 2012
*
Onsite weeks: January 30 - February 3, February 13 - 17, and February 27 -
March 2, 2012


In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its Enclosure, and your response (if any) will be available electronically for public inspection in the
The purpose of the information gathering visit is to meet with members of your staff to identify risk-significant components and operator actions. Information and documentation needed to support the inspection will also be identified. Mr. John Hanna, a Region II Senior Reactor Analyst, will accompany Mr. Sandal during the information gathering visit to review probabilistic risk assessment data and identify risk significant components which will be examined during the inspection.


NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
The enclosure lists documents that will be needed prior to the information gathering visit.
 
Sincerely,
/RA/


Rebecca Nease, Chief Engineering Branch 1 Division of Reactor Safety
Please provide the referenced information to the Region II office by January 2, 2012. Contact Mr. Sandal with any questions concerning the requested information. The inspectors will try to


Docket No.
TVA


50-390 License No. NPF-90
minimize your administrative burden by specifically identifying only those documents required for inspection preparation.


===Enclosure:===
Additional documents will be requested during the information gathering visit. The additional information will need to be made available to the team in the Region II office prior to the inspection teams preparation week of January 23, 2012. Mr. Sandal, will also discuss the following inspection support administrative details: availability of knowledgeable plant engineering and licensing personnel to serve as points of contact during the inspection; method of tracking inspector requests during the inspection; licensee computer access; working space; arrangements for site access; and other applicable information.
Inspection Report 05000390/2012008, w/Attachment: Supplemental Information


REGION II==
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Docket Nos:
050000390


License Nos: NPF-90
Thank you for your cooperation in this matter. If you have any questions regarding the information requested or the inspection, please contact Mr. Sandal at (404) 997-4513 or me at (404) 997-4519.


Report Nos:
Sincerely,
05000390/2012008
/RA/


Licensee:
Binoy B. Desai, Chief  
Tennessee Valley Authority (TVA)
 
Facility:
Watts Bar Nuclear Plant, Unit 1
 
Location:
Spring City, TN 37381
 
Dates:
January 30 - May 15, 2012
 
Inspectors:
P. Higgins, Senior Reactor Inspector (Lead)
J. Eargle, Reactor Inspector D. Mas-Penaranda, Reactor Inspector R. Lewis, Resident Inspector J. Dymek, Reactor Inspector H. Campbell, Accompanying Personnel G. Skinner, Accompanying Personnel
 
Approved by: Rebecca Nease, Chief  


Engineering Branch 1  
Engineering Branch 1  


Division of Reactor Safety
Division of Reactor Safety Docket Nos.: 50-390 License Nos.: NPF-90


=SUMMARY OF FINDINGS=
Enclosure: Information Request for Watts Bar Nuclear Plant -
IR 05000390/2012-008; 01/30/2012 - 05/15/2012; Watts Bar Nuclear Plant, Unit 1; Component


Design Bases Inspection.
Component Design Bases Inspection  


This inspection was conducted by a team of five Nuclear Regulatory Commission (NRC)inspectors from Region II, and two NRC contract personnel. Three Green non-cited violations (NCV) were identified. The significance of most findings is indicated by their color (Green,
cc w/enc/: (See page 3)  
White, Yellow, Red) using the NRC Inspection Manual Chapter (IMC) 0609, Significance Determination Process, (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, (ROP) Revision 4, dated December 2006.


NRC identified and Self-Revealing Findings
___ML112790114__________________


===Cornerstone: Mitigating Systems===
X SUNSI REVIEW COMPLETE OFFICE RII:DRS RII:DRS
*
: '''Green.'''
The team identified a non-cited violation of 10 CFR Part 50, Appendix B,
Criterion XI, Test Control, for failure to perform capacity (volumetric flow) testing on the safety-related auxiliary control air subsystem (ACAS). The licensee had documented that, for worst case environmental conditions, the air compressor capacity had little margin when compared to required air demand, even for single unit operation. This issue was entered into the licensees corrective action program as problem evaluation report 501941 for further evaluation of corrective actions.


The team determined that the failure to perform capacity testing to ensure the ACAS would meet the required air demand in response to a design basis event was a performance deficiency. This performance deficiency was more than minor because it affected the mitigating systems cornerstone attribute of equipment performance to ensure the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to develop a test procedure that would reliably ensure that the ACAS would meet required air demand for its safety-related loads during design basis accidents. The team performed a phase one significance determination process screening and determined the finding to have very low safety significance (green) utilizing the mitigating systems cornerstone column of IMC 0609, Attachment 4, Phase 1-Initial Screening and Characterization of Findings. The inspectors determined that the finding had a cross-cutting aspect in the use of conservative assumptions in the decision-making component of the human performance area. Specifically, the licensee did not use conservative assumptions in making the decision to discontinue capacity testing of the ACAS system in 2002, and stated that if that decision had been made more recently (using available internal guidance and practices regarding the testing of safety-related systems), the resulting decision would have been the same [H.1(b)]. [Section 1R21.2.4]
SIGNATURE
 
/RA/  
*
/RA/  
: '''Green.'''
The team identified a non-cited violation of 10 CFR Part 50, Appendix B,
Criterion XI, Test Control, for the licensees failure to establish a test program that demonstrated the adequacy of the auxiliary feedwater (AFW) discharge check valves. Specifically, the licensee failed to develop a test program that would provide assurance that back leakage through the AFW discharge check valves would not prevent the system from providing design flowrates to the steam generators. This issue was entered into the licensees corrective action program as problem evaluation report 499950. The licensee performed a functional evaluation and determined that the AFW system was operable based on the pumps not currently being degraded to the design limits, and the existence of additional conservatisms in the licensees design basis hydraulic analysis.
 
The team determined that the licensees failure to establish a test program to ensure that back leakage through the AFW discharge check valves would not challenge the ability of the AFW system to provide design basis flow to the steam generators was a performance deficiency. The performance deficiency was more than minor because if left uncorrected, it could have the potential to lead to a more significant safety concern. Specifically, AFW check valve back leakage could challenge the systems ability to support removal of decay heat from the reactor, which would not be identified by the licensees test program. The team performed a phase one significance determination process screening and determined the finding to have very low safety significance (green), utilizing the mitigating systems cornerstone column of IMC 0609, Attachment 4, Phase 1-Initial Screening and Characterization of Findings. Because the licensee performed a self-assessment in December 2011 that included missed opportunities to identify that check valve leakage could negatively impact the AFW system, this finding was assigned a cross-cutting aspect in the self-and independent assessments component of the problem identification and resolution area [P.3 (a)]. [Section 1R21.2.6]
 
*
: '''Green.'''
The team identified five examples of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to correctly translate vendor specifications and design calculations into maintenance and surveillance procedures. The five examples were entered into the licensees corrective action program.
 
The inspectors determined that the failure to correctly translate vendor specifications and design calculations into maintenance and surveillance procedures, was a performance deficiency. The performance deficiency was more than minor because it affected the mitigating systems cornerstone attribute of design control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In addition, the finding is similar to Inspection Manual Chapter 0612, Appendix E (example 4.a), because the failure to ensure correct translation of acceptance criteria into procedures was not isolated. The team performed a phase one significance determination process screening and determined the finding to have very low safety significance (green)utilizing the mitigating systems cornerstone column of IMC 0609, Attachment 4,
Phase 1-Initial Screening and Characterization of Findings. This finding had a cross-cutting aspect in the resources component of the human performance area, because the licensee had not ensured that complete, accurate, and up-to-date procedures were consistent with vendor and design specifications; and therefore, the procedures were not available and adequate to assure nuclear safety [H.2(c)]. [Section 1R21.2.9]
 
===Licensee-Identified Violations===
None
 
=REPORT DETAILS=
 
==REACTOR SAFETY==
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity {{a|1R21}}
 
==1R21 Component Design Bases Inspection==
{{IP sample|IP=IP 71111.21}}
 
===.1 Inspection Sample Selection Process===
The team selected risk significant components and related operator actions for review using information contained in the licensees probabilistic risk assessment. In general, this included components and operator actions that had a risk achievement worth factor greater than 1.3 or Birnbaum value greater than 1 X10-6. The sample included 15 components, including one associated with containment large early release frequency, and four operating experience items.
 
The team performed a margin assessment and a detailed review of the selected risk-significant components and operator actions to verify that the design bases had been correctly implemented and maintained. Where possible, this margin was determined by the review of the design basis and Updated Final Safety Analysis Report (UFSAR)response times associated with operator actions. This margin assessment also considered original design issues, margin reductions due to modifications, or margin reductions identified as a result of material condition issues. Equipment reliability issues were also considered in the selection of components for a detailed review. These reliability issues included items related to failed performance test results, significant corrective action, repeated maintenance, maintenance rule status, Regulatory Issue Summary 05-020 (formerly Generic Letter 91-18) conditions, NRC resident inspector input regarding problem equipment, system health reports, industry operating experience, and licensee problem equipment lists. Consideration was also given to the uniqueness and complexity of the design, operating experience, and the available defense-in-depth margins. An overall summary of the reviews performed and the specific inspection findings identified is included in the following sections of the report.
 
===.2 Component Reviews===
===.2.1 Essential Raw Cooling Water (ERCW) Discharge Check Valves (CKV-67-503F, -503H)===
====a. Inspection Scope====
The team reviewed applicable portions of the plants UFSAR, Technical Specifications (TS), system description document (SDD), and piping and instrumentation diagrams (P&IDs) to identify the design bases requirements of the ERCW check valves. The team examined system health reports, records of surveillance testing and maintenance activities, and applicable corrective actions to verify that potential degradation issues were being monitored, prevented and/or corrected. The team performed a walkdown of the ERCW system to review the installed configuration of the pumps, associated piping and check valves. Discussions with system and design engineers were conducted regarding pump and check valve design and performance requirements. A review of ERCW pump curves and inservice testing procedures was used to evaluate potential back-leakage through the subject check valves.
 
For the following operator actions, the team performed a walkthrough of associated Abnormal Operating Instructions (AOIs), Annunciator Response Instructions (ARIs),
Standard Operating Instructions (SOIs), and other applicable operations procedures (e.g. temporary instructions, etc.) with plant operators, maintenance personnel and engineers to assess operator knowledge level; adequacy of procedures; availability of special equipment when required; and the conditions under which the procedures would be performed. Detailed reviews were also conducted with operations and training department leadership to further understand and assess the procedural rationale and approach to meeting the design basis and UFSAR response and performance requirements. Selected operator actions associated with the following events/evolutions were reviewed:
* Crosstie opposite unit and train ERCW headers (1B-B strainer plugged) -
[HAERCW1A]
* Operators fail to clear ERCW screens before plant trip - [DHAERCWS]
 
====b. Findings====
No findings were identified.
 
===.2.2 Residual Heat Removal (RHR) Pump (1A-A, 1PMP-074-0010-A)===
====a. Inspection Scope====
The team reviewed the UFSAR, system design criteria, current system health report, selected drawings, operating procedures, and past corrective action reports for the RHR pump. The team interviewed the RHR system engineer to discuss the overall health and condition of the RHR system and pumps. Further, the team interviewed inservice test engineers responsible for maintaining and updating the required TS surveillance procedures. Also, the team interviewed instrumentation and control engineers responsible for performing the associated instrumentation uncertainty calculations.
 
Calculations addressing required RHR pump performance requirements during design bases accidents, calculations addressing the uncertainties of the instruments used to verify pump performance during required inservice test surveillances, quarterly and comprehensive surveillance procedures, and test results used to verify required RHR pump performance were reviewed. A sample of emergency operating procedures, which incorporated RHR pump, flow instrumentation and start logic were also reviewed.
 
The team verified by review of schematic drawings, that operation of the pump motors was consistent with the design basis and operational requirements. The team reviewed the protection coordination calculation for the RHR pump motor and verified that the circuit breaker ratings and protective devices trip settings and alarm functions were consistent with the licensing basis and operational requirements. The team verified that the brake horsepower required by the pump was within the motor rating. The team performed a walkdown of the RHR system to assess observable material condition of the pump motors. The team verified that the ambient conditions were consistent with vendor recommendation for the motors.
 
====b. Findings====
No findings were identified.
 
===.2.3 Standby Diesel Generator (SDG) Heat Exchangers (2B1 and 2B2)===
====a. Inspection Scope====
The team reviewed the plant TS, UFSAR, SDDs and associated P&IDs to establish an overall understanding of the cooling water requirements for the SDG systems. Design calculations were reviewed to verify that the design basis and design assumptions had been appropriately translated into these documents. The team also reviewed 10 CFR Appendix A to Part 50; General Design Criteria (GDC) 44 Cooling Water, GDC 45, Inspection of Cooling Water and GDC 46 Testing of Cooling Water as well as Generic Letter 89-13 Service Water System Problems Affecting Safety-Related Equipment.
 
Component walkdowns were performed to verify the equipment was installed in accordance with design drawings, system descriptions and that the system was being maintained in accordance with the design basis. The overall condition of areas containing this equipment was examined, and that cooling water flow control and throttling valves to the heat exchangers were in their correct positions and properly secured. Test procedures and results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses served to validate component operation under accident/event conditions. Heat exchanger vendor documentation was reviewed to verify assumed flow rates, temperature differential(s)and fouling factor limits were accounted for in plant technical instructions and surveillances. System health reports, preventative and corrective maintenance, inservice testing and surveillance testing was reviewed to confirm that testing is being done in accordance with manufacturers requirements. Docketed plant correspondence was reviewed to determine plant commitments to Generic Letter 89-13 for initial baseline testing and periodic re-testing of open cycle service water systems against minimum frequency intervals as specified in Generic Letter 89-13.
 
====b. Findings====
No findings were identified.
 
===.2.4 Auxiliary Control Air Subsystem (ACAS) (A-A and B-B)===
====a. Inspection Scope====
The team reviewed applicable portions of the plants TS, UFSAR, SDD, and P&IDs to establish an overall understanding of the design bases of the safety-related ACAS.
 
Design calculations were reviewed to verify that design basis air capacity flow rate requirements had been appropriately translated into these documents. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented. Also, current surveillance testing including start-logic, and previous testing, which included compressor air capacity testing, was reviewed. Further, several discussions with both the current and former system engineers took place, with focus on maintenance practices and testing requirements and results of the ACAS.
 
For the following operator actions, the team performed a walkthrough of associated AOIs, ARIs, SOIs, and other operations procedures (e.g. temporary instructions, etc.) with plant operators, maintenance personnel and engineers to assess operator knowledge level; adequacy of procedures; availability of special equipment when required; and the conditions under which the procedures would be performed. Detailed reviews were also conducted with operations and training department leadership to further understand and assess the procedural rationale and approach to meeting the design basis and UFSAR response and performance requirements. Selected operator actions associated with the following event/evolution were reviewed:
* Restore Auxiliary Feedwater (AFW) control following initiator and loss of air [HAFR1]
 
====b. Findings====
=====Introduction:=====
The team identified a green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to perform capacity (volumetric flow) testing on the safety-related auxiliary control air subsystem (ACAS). The licensee had documented that for worst case environmental conditions, the compressor capacity had little margin when compared to required air demand, even for single unit operation.
 
=====Description:=====
The ACAS is a safety-related system required to supply air to safety-related equipment during design basis events (e.g., air is to be supplied at adequate pressure and capacity to auxiliary feedwater system level control valves and steam generator pressure control valves during postulated design basis accidents). Although the licensees ACAS test procedures demonstrated start logic functionality and the capability to develop adequate system pressure, the test procedures did not evaluate or measure compressor air capacity. Capacity testing was previously performed on this system by the licensee, but was removed from the test program in 2002. Instead of capacity testing, the licensee relied upon the preventive maintenance program, during which various component parts were replaced, to ensure the system remained capable of operating at proper capacity. The licensee did not perform as-found capacity tests prior to such maintenance and did not perform post-maintenance capacity testing. The licensee had concluded that under design basis conditions, the ACAS capacity margin was less than 5 standard cubic feet per minute, and that this represents little margin (even for single unit operation). In response to the teams questions, this issue was entered into the licensees corrective action program as problem evaluation report 501941.
 
=====Analysis:=====
The team determined that the failure to perform capacity testing to ensure the ACAS would meet the required air demand in response to a design basis event was a performance deficiency. This performance deficiency was more than minor because it affected the mitigating systems cornerstone attribute of equipment performance to ensure the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to develop a test procedure that would reliably ensure that the ACAS would meet required air demand for its safety-related loads during design basis accidents. The team performed a phase one significance determination process screening and determined the finding to have very low safety significance (green)utilizing the mitigating systems cornerstone column of Inspection Manual Chapter 0609, Attachment 4, Phase 1-Initial Screening and Characterization of Findings. The finding did not represent a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the finding had a cross-cutting aspect in the use of conservative assumptions in the decision-making component of the human performance area. Specifically, the licensee did not use conservative assumptions in making the decision to discontinue capacity testing of the ACAS system in 2002, and stated that if that decision had been made more recently (using available internal guidance and practices regarding the testing of safety-related systems), the resulting decision would have been the same
[H.1(b)].


=====Enforcement:=====
NAME S. SANDAL B. DESAI
10 CFR Part 50, Appendix B, Criterion XI, Test Control, states in part, that a program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate acceptance limits contained in applicable documents. Contrary to this requirement, since 2002, the licensee had failed to establish a test control program to assure that the ACAS would perform satisfactorily in response to a design basis event. Specifically, the licensee did not have a test procedure to verify that ACAS capacity met design basis accident performance requirements. Because this violation was of very low safety significance and was entered into the licensees corrective action program, this violation is being treated as an NCV, consistent with the Enforcement Policy: NCV 05000390/2012008-01, Failure to Establish Test Procedures to Assure Satisfactory ACAS Performance during Design Basis Accidents.


===.2.5 RHR Pump Room Coolers (A-A and B-B)===
DATE 10/ 03 /2011 10/ 06 /2011
====a. Inspection Scope====
The team reviewed applicable portions of the plants TS, UFSAR, SDDs, and P&IDs to establish an overall understanding of the design bases of the coolers. Design calculations were reviewed to verify that design basis heat removal requirements, capability, and flow rates had been appropriately translated into these documents. Test procedures and results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents, and that individual tests and/or analyses served to validate component operation under accident/event conditions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with inservice/equipment qualification life. Environmental qualification and seismic documents/calculations were reviewed to verify that the cooler was appropriately qualified. Component walkdowns were conducted to verify that the installed configurations would support their design basis function under accident conditions and had been maintained to be consistent with design assumptions.


====b. Findings====
E-MAIL COPY?
No findings were identified.
YES NO YES NO YES NO YES NO YES NO YES NO YES NO


===.2.6 Auxiliary Feedwater Check Valve===
TVA
(3-805)


====a. Inspection Scope====
cc w/encl:
The team reviewed applicable portions of the plants TS, UFSAR, SDDs, and P&IDs to establish an overall understanding of the design bases of the check valve. Design calculations (e.g., differential pressure) were reviewed to verify that the design basis and design assumptions had been appropriately translated into these documents. Operating procedures were reviewed to verify that operator actions involving the high-pressure fire protection system alignment of auxiliary feedwater (AFW) were feasible. Test procedures and results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses served to validate component operation under accident/event conditions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with inservice/equipment qualification life. Seismic documents/calculations were reviewed to verify that the valve was appropriately qualified. Component walkdowns were conducted to verify that the installed configurations would support their design basis function under accident conditions and had been maintained to be consistent with design assumptions.


====b. Findings====
D. E. Grissette Site Vice President Watts Bar Nuclear Plant Tennessee Valley Authority Electronic Mail Distribution
=====Introduction:=====
The team identified a green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to establish a test program that demonstrated the adequacy of the AFW discharge check valves. Specifically, the licensee failed to develop a test program that would provide assurance that back leakage through the AFW discharge check valves would not prevent the system from providing design flowrates to the steam generators.


=====Description:=====
G. A. Boerschig Plant Manager Watts Bar Nuclear Plant, MOB 2R-WBN Tennessee Valley Authority Electronic Mail Distribution
The AFW system supplies feedwater to the steam generators in the event of a loss of main feedwater to remove reactor decay heat and avoid reactor coolant system over pressurization. The AFW system consists of two motor-driven pumps and one turbine-driven pump. Check valves are installed on the discharge of the pumps with design functions of
: (1) opening on an AFW initiation, and
: (2) closing on a pump failure to maintain the running pumps flow path integrity. If a check valve or combination of check valves on a non-running pump were to leak, that flow would be re-circulated back to the condensate storage tank rather than being delivered to the steam generators.


The team determined that calculation HCGTBG091981, Design Parameter for Motor and Turbine Driven AFW Pumps, Rev. 8, did not account for any check valve leakage losses.
C. J. Riedl Acting Manager, Licensing Watts Bar Nuclear Plant, ADM 1L-WBN Tennessee Valley Authority P.O. Box 2000 Spring City, TN 37381


The team noted that at worst case allowable degradation of the pumps, little to no margin existed in meeting the system flow requirements for various system operating conditions.
J. W. Shea Manager, Corp. Nuclear Licensing - WBN Tennessee Valley Authority Electronic Mail Distribution


In this regard, under the worst case scenario of a loss of the turbine-driven AFW pump and a 5% degradation of both motor-driven AFW pumps (as accounted for in the licensees hydraulic analysis), the allowable loss due to check valve back leakage would be zero. The team determined that licensee has no testing provisions in place to ensure that the zero leakage condition was met for the configuration described above, or other possible system alignments.
E. J. Vigluicci Assistant General Counsel Tennessee Valley Authority Electronic Mail Distribution


This issue was entered into the licensees corrective action program as problem evaluation report 499950. The licensee performed a functional evaluation and determined that the AFW system was operable based on the pumps performance not currently being degraded to the design limit, and the existence of additional conservatisms in the licensees hydraulic analysis. Additionally, the team noted that the licensee had performed a self-assessment in December 2011, to determine if check valve back leakage could challenge design basis analyses used to demonstrate the adequacy of system design. Although the licensee evaluated the essential raw cooling water system check valves for this potential vulnerability, the licensee had not applied the evaluation to other safety-related systems including AFW.
W. D. Crouch Licensing Manager, Unit 2 Watts Bar Nuclear Plant, EQB 1B-WBN Tennessee Valley Authority P.O. Box 2000 Spring City, TN 37381


=====Analysis:=====
County Mayor P.O. Box 156 Decatur, TN 37322
The team determined that the licensees failure to establish a test program to ensure that back leakage through the AFW discharge check valves would not challenge the ability of the AFW system to provide design basis flow to the steam generators, was a performance deficiency. The performance deficiency was more than minor because if left uncorrected, the performance deficiency could have the potential to lead to a more significant safety concern. Specifically, AFW check valve back leakage could challenge the systems ability to support removal of decay heat from the reactor and this condition would not be identified by the licensees test program. The team performed a phase one significance determination process screening and determined the finding to have very low safety significance (green) utilizing the mitigating systems cornerstone column of IMC 0609, Attachment 4, Phase 1-Initial Screening and Characterization of Findings. The finding did not represent a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. Because the licensee performed a self-assessment in December 2011, that included missed opportunities to identify that check valve leakage could negatively impact the AFW system, this finding was assigned a cross-cutting aspect in the self-and independent assessments component of the problem identification and resolution area [P.3(a)].


=====Enforcement:=====
County Executive 375 Church Street Suite 215 Dayton, TN 37321
Appendix B to 10 CFR Part 50, Criterion XI states in part, that a test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Contrary to this requirement, since initial plant operation on February 7, 1996, the licensee failed to establish a test program that demonstrated that the AFW discharge check valves would perform satisfactorily in service. Specifically, the licensee failed to develop a test program that would provide assurance that back leakage through the AFW discharge check valves would not prevent the system from providing design flowrates to the steam generators. Because this violation was of very low safety significance and it was entered into the licensees corrective action program, this violation is being treated as an NCV, consistent with the Enforcement Policy: NCV 05000390/2012008-02, Failure to Adequately Test the AFW Discharge Check Valves.


===.2.7 Component Cooling Water Heat Exchangers===
Tennessee Department of Environment &
====a. Inspection Scope====
Conservation Division of Radiological Health 401 Church Street Nashville, TN 37243
The team reviewed applicable portions of the plants TS, UFSAR, SDDs, and P&IDs to establish an overall understanding of the design bases of the heat exchangers. Design calculations were reviewed to verify that design basis heat removal requirements, capability, and flow rates had been appropriately translated into these documents. Test procedures and results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents, and that individual tests and/or analyses served to validate component operation under accident/event conditions. Control panel indicators were observed and operating procedures reviewed to verify that component operation and alignments were consistent with design and licensing basis assumptions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored, or prevented, and the component replacement was consistent with inservice/equipment qualification life. Component walkdowns were conducted to verify that the installed configurations would support their design basis function under accident conditions and had been maintained to be consistent with design assumptions.


For the following operator actions, the team performed a walkthrough of associated AOIs, ARIs, SOIs, and other operations procedures (e.g., temporary instructions, etc.) with plant operators, maintenance personnel, and engineers to assess operator knowledge level; adequacy of procedures; availability of special equipment when required; and the conditions under which the procedures would be performed. Detailed reviews were also conducted with operations and training department leadership to further understand and assess the procedural rationale and approach to meeting the design basis and UFSAR response and performance requirements. Selected operator actions associated with the following event/evolution were reviewed:
Senior Resident Inspector U.S. Nuclear Regulatory Commission Watts Bar Nuclear Plant U.S. Nuclear Regulatory Commission 1260 Nuclear Plant Road Spring City, TN 37381-2000
* Align and initiate alternate cooling to centrifugal charging pump A [HACCSR2]


====b. Findings====
Ann Harris 341 Swing Loop Rockwood, TN 37854
No findings were identified.


===.2.8 Auxiliary Feedwater Pressure Control Valve (3-132)===
Enclosure INFORMATION REQUEST FOR WATTS BAR NUCLEAR PLANT COMPONENT DESIGN BASES INSPECTION
====a. Inspection Scope====
The team reviewed applicable portions of the plants TS, UFSAR, SDDs, and P&IDs to establish an overall understanding of the design bases of the valve. Design calculations (i.e., differential pressure and required torque/thrust) were reviewed to verify that the design basis and design assumptions had been appropriately translated into these documents. Test procedures and results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses served to validate component operation under accident/event conditions. Control panel indicators were observed and operating procedures reviewed to verify that component operation and alignments were consistent with design and licensing basis assumptions. The team also reviewed instrument loop diagrams, loop setpoints and scaling documents, and loop accuracy calculations to verify that the valve controls were consistent with design bases assumptions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with inservice/equipment qualification life. Component walkdowns were conducted to verify that the installed configurations would support their design basis function under accident conditions and had been maintained to be consistent with design assumptions.


====b. Findings====
Please provide the information electronically in.pdf files, Excel, or other searchable format on CDROM (or FTP site, Sharepoint, etc.) The CDROM (or website) should be indexed and hyperlinked to facilitate ease of use.
No findings were identified.


===.2.9 Shutdown Board (1B-B)===
1.
====a. Inspection Scope====
The team reviewed bus loading calculations to determine whether the 6.9 kilovolt (kV)system had sufficient capacity to support its required loads under worst case accident loading and grid voltage conditions. The team reviewed the design of the degraded voltage protection scheme to determine whether it afforded adequate voltage to safety related devices at all voltage distribution levels. This included review of degraded voltage relay setpoint calculations, review of the degraded voltage logic scheme, and the licensees response to NRC Information Notice 95-05, Undervoltage Protection Relay Settings Out of Tolerance Due to Test Equipment Harmonics. The team reviewed the overcurrent protection scheme for the 6.9 kV buses including drawings and calculations to determine whether loads were adequately protected and immune from spurious tripping. The team reviewed 125 volt direct current (VDC) system voltage drop calculations to determine whether 6.9 kV bus circuit breakers had adequate control voltage. The team reviewed the load shedding and load sequencing schemes to determine whether they were consistent with the design bases and design calculations.


The team reviewed maintenance schedules and procedures for the 6.9 kV bus and its associated circuit breakers to determine whether the equipment was being properly maintained. This included reviewing acceptance criteria in procedures for consistency with vendor recommendations and design calculations. The team reviewed the fast bus transfer scheme, including drawings, calculations and procedures, to determine whether the transfer capability described in the UFSAR could be achieved without adverse effects on equipment and systems. The team reviewed calculations for switchgear temperature rise to determine to assess the effect of a loss of ventilation. The team reviewed corrective action documents and maintenance records to determine whether there were any adverse operating trends. In addition, the team performed a visual inspection of the 6.9 kV safety buses to assess material condition and the presence of hazards.
From your most-recent Probabilistic Safety Analysis (PSA) excluding external events and fires:


For the following operator actions, the team performed a walkthrough of associated AOIs, ARIs, SOIs, and other operations procedures (e.g., temporary instructions, etc.) with plant operators, maintenance personnel and engineers to assess operator knowledge level; adequacy of procedures; availability of special equipment when required; and the conditions under which the procedures would be performed. Detailed reviews were also conducted with operations and training department leadership to further understand and assess the procedural rationale and approach to meeting the design basis and UFSAR response and performance requirements. Selected operator actions associated with the following event/evolution were reviewed:
a. Two risk rankings of components from your site-specific probabilistic safety analysis (PSA): one sorted by Risk Achievement Worth (RAW), and the other sorted by Birnbaum Importance
* Operator action to crosstie shutdown boards [HASBXT]


b.
b. A list of the top 500 cutsets


===.1 Findings and Observations===
2.
=====Introduction:=====
The team identified a green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, with five examples, for the licensees failure to correctly translate vendor specifications and design calculations into maintenance and surveillance procedures.


=====Description:=====
From your most-recent Probabilistic Safety Analysis (PSA) including external events and fires:  
The team identified the following instances where the design as documented in calculations and vendor instructions was not properly translated into maintenance procedures:  


6.9 kV and 480 volt (V) Breaker Operating Voltage Maintenance procedure MI-57.002 provided instructions for checking the operating voltage of 480 V DS type circuit breaker close coils at reduced voltage. A note associated with the applicable step stated that if the minimum voltage exceeds 100 VDC then a problem evaluation report (PER) may be required. The minimum close coil voltage specified by the manufacturer and used as acceptance criteria in calculation WBN EEB-MS-TI11-0004 was 90 VDC. Consequently, an operating voltage of 100 VDC may not be adequate to ensure proper breaker operation under design basis conditions at the start of an accident. A similar discrepancy existed in maintenance procedure MI-57.001 for 6.9 kV Magne-blast circuit breakers, which also cited an acceptance criteria of 100 VDC versus the 90 VDC used in the design basis. In response to this concern, the licensee initiated PER 507645, and performed a partial review (approximately 85%) of maintenance records to confirm that no test results for installed breakers had failed to meet the manufacturers rating of 90 VDC.
a. Two risk rankings of components from your site-specific Probabilistic Safety Analysis (PSA): one sorted by Risk Achievement Worth (RAW), and the other sorted by Birnbaum Importance


4160 V to 480 V Transformer Insulation Resistance Maintenance procedure MI-57.200 provided acceptance criteria for the 4160 V to 480 V transformer insulation resistance (IR) as 1.5 mega-ohm secondary and 30 mega-ohm primary. Vendor manual VTD W120 2356 specified the minimum IR as 32 mega-ohm for 1.2 kV class windings and 230 mega-ohm for 8.7 kV insulation class windings. In response to this concern, the licensee issued PER 511469 and confirmed that the last performance of maintenance procedure MI-57.200 for each 480 V shutdown board transformer showed actual IR readings above the manufacturers minimum recommendations.
b. A list of the top 500 cutsets


Breaker Closing Time Maintenance procedure MI-57.001, dated July 5, 2011, provided for as-left testing for the closing time of 6.9kV circuit breakers after maintenance and provided acceptance criterion of less than or equal to 84 milliseconds (5 cycles). Calculation WBN EEB-MS-TI06-0027 evaluated the effects of the fast transfer capability of the 6.9 kV systems and determined that a dead bus time of approximately 4.1 cycles could result in exceeding the industry standard evaluation criteria of 1.33 Volt/Hertz. The acceptance criteria in procedure MI-57.001 was not consistent with the maximum desirable dead bus time determined in the calculation. In addition, maintenance procedure MI-57.001, Step 6.2
3.
[23] did not provide any acceptance criteria for as-found closing time testing. The team noted that the last test for the alternate source breaker for 6.9 kV bus 1B-B, was 72 milliseconds, which was consistent with the dead bus time evaluated in calculation WBN EEB-MS-TI06-0027. In response to this concern, the licensee issued PER 517095.


480 V Breaker Service Life Limitations The vendor manual for Westinghouse type DS circuit breakers specified breaker and subcomponent service life limits. Maintenance procedure MI-57.002 identified the service life limit for the breaker (either 4000 or 1500 cycles, depending on model), but did not reference the more limiting service life limits for breaker subcomponents such as the direct trip actuator (400 cycles) and overcurrent trip switch (400 cycles). In response to this concern the licensee initiated PER 511010 and provided data to justify that no subcomponents have exceeded their service life limits.
Risk ranking of operator actions from your site specific PSA sorted by RAW. Provide human reliability worksheets for these items.


Specification for Harmonic Distortion Limits for Relay Test Sets Vendor manual WBN-VTD-AS04-0080 for Asea Brown Boveri (ABB) type 27N degraded voltage relays specified the use of a calibration voltage source with a maximum harmonic distortion of 0.3%. Surveillance procedure 1-SI-211-5-B, Step 4.2 [2] specified the use of a Doble relay test set with a maximum harmonic distortion of 2%, which was non-conservative with respect to the vendor specification. In response to this concern, the licensee issued PER 513893 and confirmed that the Doble test sets actually available and used for calibrating the degraded voltage relays conformed to the vendors specifications.
4.


=====Analysis:=====
List of time critical operator actions with a brief description of each action.
The inspectors determined that the failure to correctly translate vendor specifications and design calculations into maintenance and surveillance procedures was a performance deficiency. The performance deficiency was more than minor because it affected the mitigating systems cornerstone attribute of design control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In addition, the finding is similar to Inspection Manual Chapter 0612, Appendix E (example 4.a), because the failure to ensure correct translation of acceptance criteria into procedures was not isolated. The team performed a phase one significance determination process screening and determined the finding to have very low safety significance (green) utilizing the mitigating systems cornerstone column of IMC 0609, Attachment 4, Phase 1-Initial Screening and Characterization of Findings. The finding did not represent a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the resources component of the human performance area, because the licensee had not ensured that complete, accurate, and up-to-date procedures were consistent with vendor and design specifications; and therefore, the procedures were not available and adequate to assure nuclear safety [H.2(c)].  


=====Enforcement:=====
5.
Appendix B to 10 CFR 50, Criterion III, Design Control, requires in part, that measures be established to assure that the design basis is correctly translated into specifications, drawings, procedures, and instructions. Contrary to this requirement, since the revision of maintenance procedure MI-57.001 on July 5, 2011, the licensees design control measures had failed to assure that vendor specifications and design calculations were correctly translated into specifications relied on in maintenance and surveillance procedures. Specifically, the licensee failed to implement measures to ensure that specifications for circuit breaker control voltage, transformer insulation resistance, circuit breaker closing times, circuit breaker service life limitations, and relay test set harmonic distortion were correctly reflected in maintenance and surveillance procedures. Because this violation was of very low safety significance and it was entered into the licensees corrective action program, this violation is being treated as an NCV, consistent with the Enforcement Policy: NCV 05000390/2012008-03, Inadequate Acceptance Criteria in Maintenance and Surveillance Procedures (5 Examples).


b.
List of Emergency and Abnormal Operating Procedures revised (significant) since April 1, 2010 with a brief description of each revision.


===.2 Findings and Observations===
6.
=====Introduction:=====
The team identified an issue of concern and an unresolved item related to the effect of electrical system harmonics on safety-related degraded voltage relays.


Specifically, in 1993, the licensee identified that harmonic distortions adversely affected the 6.9 kilovolt (kV) bus overvoltage relays by causing them to alarm unnecessarily. The licensee entered this issue into their corrective action program and modified the overvoltage relays to minimize the effects. However, the licensee did not identify (or otherwise evaluate) the adverse effect that harmonics could have on the ability of the degraded voltage relays to perform their safety function, as required by limiting condition for operation 3.3.5 of the plants technical specifications.
List of components with low design margins (i.e., pumps closest to the design limit for flow or pressure, diesel generator close to design required output, heat exchangers close to rated design heat removal, MOV risk-margin rankings, etc.) and associated evaluations or calculations.


=====Description:=====
7. List of station operating experience evaluations/reviews performed and documented in the stations corrective action program for industry events and safety related equipment failures/vulnerabilities [as communicated by NRC generic communications, industry communications, 10 CFR part 21 notifications, etc.] since April 1, 2010.
The Watts Bar degraded voltage protection scheme features three ABB type 27N relays for each 6.9 kV safety bus, arranged in a two out of three tripping scheme. The ABB instruction bulletin 7.4.1.7-7 contained in vendor manual WBN-VTD-AS04-0080 states that
: (1) the relay employs a peak voltage detector, and
: (2) harmonic distortion on the AC waveform can have a noticeable effect on the relay operating point and the measuring instruments used to calibrate the relay. The bulletin also notes that the relay is available with an internal harmonic filter for applications where waveform distortion is a factor. The team noted that calculation WBPE2119202001, 6.9kV Shutdown & Logic Boards Undervoltage Relay Requirements/Demonstrated Accuracy Calculation, identified the relay as a model not equipped with a harmonic filter, but did not address the basis for excluding harmonic distortion as a factor which affected relay accuracy. In response to the teams inquiries, the licensee provided PER 930397 that addressed spurious actuations of the ABB type 59H overvoltage relays which are similar in design to the ABB type 27N degraded voltage relays. Troubleshooting tests performed to identify the cause of the 59H spurious actuations revealed that high levels of 6.9 kV system harmonics from sources both external and internal to the station accompanied the spurious operations. The causal factor section of PER 930397 stated that the relays sometimes trip on harmonic distortion although the root mean square voltages are at acceptable levels. Corrective actions consisted of replacing the type 59H overvoltage relays with a model equipped with harmonic filters. The team further noted that the extent of condition section of PER 930397 did not identify or address whether the degraded voltage relays operating point could also be affected by the same harmonics implicated in the maloperation of the overvoltage relays.


The team was concerned that harmonics on the 6.9 kV system could cause the degraded voltage relays to fail to actuate at the setpoint specified by technical specifications. Persistent harmonics can be produced by factors external to the nuclear site or by internal phenomena. A typical internal source of harmonics at nuclear power plants is motor defects. The team was also concerned that transient harmonics could cause the relays to spuriously reset during an actual degraded voltage event, thereby delaying the protective function beyond the 10 seconds stipulated in technical specification limiting condition for operation 3.3.5. Specifically, the degraded voltage relays design features an instantaneous reset characteristic that could allow reset of the degraded voltage relay in less than two cycles in the presence of harmonics, thereby reinitiating the external 10 seconds timer. The reset function of the existing degraded voltage relays is identical to the tripping function of the overvoltage relays that actuated due to transient harmonics in 1993. In 1993, transient harmonics were measured at levels of greater than 10% total harmonic distortion during the troubleshooting for PER 930397 versus the 0.3% distortion deemed acceptable by the relay vendor. The transient harmonics documented in PER 930397 were attributed to events that included the trip of the nearby Sequoyah generating station, and to breaker operations at the Watts Bar station. The team noted that similar conditions could exist during an accident scenario when proper performance of the degraded voltage scheme time delay would be critical with respect to satisfying the response time assumptions in the accident analysis.
8.


In response to the teams concerns, the licensee provided information regarding condition monitoring of large motors that consisted of periodic measurement and analysis of motor bearing vibration from which various defects that may produce harmonics could be identified. The team noted, however, that there was no written guidance or acceptance criteria for these tests that would prompt engineering to investigate whether suspected motor defects could produce harmonics that would adversely affect the accuracy of degraded voltage relays. Specifically, there was no recognition in design or maintenance documents regarding the susceptibility of the degraded voltage relays to harmonic distortion, or the need to investigate suspected motor defects with respect to this susceptibility. The team further noted that during normal bus voltage conditions when voltage is above the degraded voltage relay reset setpoint, harmonics would shift system peak voltage away from the degraded voltage relay operating setpoint rather than closer to it, and so the presence of harmful harmonics would not self-reveal by spurious actuations. The overvoltage relays are now equipped with harmonic filters so they will also not reveal the presence of either transient or persistent harmonics. Based on the teams observations, the licensee has entered these concerns into their corrective action program as PER 515413 and PER 546072.
List and brief description of safety related SSC design modifications implemented since April 1, 2010.


Summary: The team determined that additional review of information recently received from the licensee regarding Watts Bars design and licensing bases was necessary to determine if the licensees performance constituted a violation of NRC regulatory requirements. Additionally, the team determined that additional consultation with the Office of Nuclear Reactor Regulation was warranted before reaching a final disposition of the unresolved item. This unresolved item is open pending
9.
: (1) the review of additional information from the licensee regarding the design and licensing basis of the degraded voltage relays and
: (2) consultation with the Office of Nuclear Reactor Regulation: URI 05000390/2012008-04, Effect of System Harmonics on Degraded Voltage Relay Function.


===.2.10 Shutdown Transformer (1A1-A)===
List and brief description of common-cause component failures that have occurred since April 1, 2010.
====a. Inspection Scope====
The team reviewed load flow calculations to determine whether the transformer was applied within its specified ratings. The team reviewed maintenance schedules, vendor recommendations, and procedures to determine whether the transformers were being properly maintained. This included reviewing acceptance criteria in procedures for consistency with vendor recommendations and design calculations. The team reviewed protective relaying schemes and calculations to determine whether the transformer was adequately protected and whether it was susceptible to spurious tripping. The team reviewed maintenance and corrective action histories to determine whether there have been any adverse operating trends. In addition, the team performed a walkdown of the installed equipment to determine whether the installed configuration is consistent with design documents including drawings, and calculations, and to assess the presence of hazards.


====b. Findings====
Enclosure 10. List and brief description of operability evaluations completed since April 1, 2010.
No findings were identified.


===.2.11 Component Cooling System Pump Motor (1A-A)===
11. List of equipment on the sites Station Equipment Reliability Issues List, including a description of the reason(s) why each component is on that list and summaries (if available)
====a. Inspection Scope====
of your plans to address the issue(s).
The team reviewed maintenance schedules, and procedures, and completed work orders to determine whether the motor was being properly maintained. The team review protective relaying schemes and calculations to determine whether the motor was adequately protected and whether it was susceptible to spurious tripping. The team reviewed maintenance and corrective action histories to determine whether there have been any adverse operating trends. The team also reviewed the component cooling system pump 1A-A motor circuit breaker. Corrective action and maintenance records were reviewed to determine if there had been any adverse operating trends. In addition, the team performed a walkdown of the installed equipment to determine whether the installed configuration is consistent with design documents including drawings, and calculations, and to assess the presence of hazards.


====b. Findings====
12. List and brief description of equipment currently in degraded or nonconforming status as described in RIS 05-020.
No findings were identified.


===.2.12 Auxiliary Feedwater Level Control Valves (3-156, -164)===
13. List and reason for equipment classified in maintenance rule (a)(1) status since April 1, 2010 to present.
====a. Inspection Scope====
The TS, UFSAR, SDs, and P&IDs, were reviewed to establish an overall understanding of the design bases of the valves. Design calculations (e.g., differential pressure and required torque/thrust) were reviewed to verify that the design basis and design assumptions had been appropriately translated into these documents. Test procedures and results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses served to validate component operation under accident/event conditions. Control panel indicators were observed and operating procedures reviewed to verify that component operation and alignments were consistent with design and licensing basis assumptions. The team also reviewed instrument loop diagrams, loop setpoints and scaling documents, and loop accuracy calculations to verify that the valve controls were consistent with design bases assumptions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with inservice/equipment qualification life. Component walkdowns were conducted to verify that the installed configurations would support their design basis function under accident conditions and had been maintained to be consistent with design assumptions.


====b. Findings====
14. Copies of System Descriptions (or the like design basis documents) for Safety-Related Systems.
No findings were identified.


===.2.13 Refueling Water Storage Tank LCVs (62-135A, -136B)===
15. Copy of UFSAR(s).
====a. Inspection Scope====
The team reviewed applicable portions of the plants TS, UFSAR and system descriptions to identify design basis requirements for LCV-62-135A/136B. The team reviewed the calculations that establish control circuit voltage drop, short circuit, and protection/coordination including thermal overload sizing and application to verify adequate protection during design bases scenarios. The team verified by review of control diagrams, that the operation of the LCV was consistent with the design basis and operational requirements. The team interviewed the system engineer to discuss the valve analysis as well as operational and maintenance history to verify that potentially degraded conditions were being appropriately addressed. Test procedures and recent test results were reviewed against design bases documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and analyses served to validate component operation under accident conditions. The team examined maintenance rule documentation to verify that the valves were properly scoped, and monitored. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored, or prevented, and that scheduled component replacements were consistent with vendor recommendations and equipment qualification life. Component walkdown was conducted to verify that the installed configurations would support the design basis function under accident conditions and had been maintained to be consistent with design assumptions. The team conducted a non-intrusive visual inspection of LCV-62-135A/136B to verify that any potentially degraded material conditions were being appropriately addressed. Also, the team verified testing and calibration of instruments related to the valve.


====b. Findings====
16. Copy of Technical Specification(s).
No findings were identified.


===.2.14 125 Volt Battery Boards (I and II)===
17. Copy of Technical Specifications Bases.
====a. Inspection Scope====
The team reviewed the design basis documentation and UFSAR to identify the loading requirements for the vital batteries. The team reviewed the inputs to the battery sizing analysis and the battery voltage study, TS and maintenance allowable terminal load resistance, and panel load schedules to verify the adequate sizing of the battery. The battery voltage study was reviewed to verify adequate voltage was available to critical components. The vendor manual was reviewed to verify battery installation and operating instructions were implemented. Battery TS surveillance test and inspection results were reviewed to verify degradation was identified and anomalies were addressed and corrected. The equipment history as indicated by corrective work orders and condition reports was reviewed to verify that identified equipment problems were corrected. A field walkdown was performed to assess observable material conditions of the batteries.


Also, the team reviewed schematic diagrams and calculations for the normal supply breakers to determine whether equipment operation was consistent with the design bases. The team reviewed calculations for protective device settings to determine whether the breakers were subject to spurious tripping, and whether the breakers were selectively coordinated with upstream devices. Also, the team reviewed associated corrective action history to verify that degraded conditions were being appropriately addressed. In addition, the team interviewed the system engineer and performed a non-intrusive visual inspection of the direct current bus to assess the installation configuration and verify that degraded material conditions were being appropriately addressed.
18. Copy of Technical Requirements Manual(s).


====b. Findings====
19. List and brief description of Root Cause Evaluations that have been performed since
No findings were identified.


===.2.15 Hydrogen Igniters [large early release frequency]===
April 1, 2010.
====a. Inspection Scope====
The team reviewed applicable portions of the plants TS, UFSAR and system descriptions to identify design basis requirements for the hydrogen igniters. The team interviewed the system engineer to discuss operation and maintenance history to verify that potentially degraded conditions were being appropriately addressed. Operation procedures for emergency power to the hydrogen igniters were reviewed to verify that component operation and power supply alignment were consistent with the design. Test procedures and recent test results were reviewed against design bases documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and analyses served to validate component operation. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and that scheduled component replacements were consistent with vendor recommendations and equipment qualification life.


====b. Findings====
20. In-service Testing Program Procedure(s).
No findings were identified.


===.3 Operating Experience===
21. Corrective Action Program Procedure(s).
====a. Inspection Scope====
The team reviewed four operating experience issues for applicability at Watts Bar Nuclear Plant. The team performed an independent review for these issues and where applicable, assessed the licensees evaluation and dispositioning of each item. The issues that received a detailed review by the team included:
* NRC Generic Letter 1988-14, Instrument Air Supply System Problems Affecting Safety-Related Equipment
* NRC Regulatory Information Summary 2000-05, Spring Actuated Safety and Relief Valve Reliability
* NRC Information Notice 1992-29, Potential Breaker Miscoordination Caused by Instantaneous Trip Circuitry
* NRC Information Notice 1987-08, Degraded Motor Leads in LIMITORQUE DC Motor Operators


====b. Findings====
22. One line diagram of electrical plant (electronic and full size - hard copy week of January 9).
No findings were identified.


==OTHER ACTIVITIES==
23. Index and legend for electrical plant one-line diagrams.
{{a|4OA6}}


==4OA6 Meetings, Including Exit==
24. Primary AC calculation(s) for safety-related buses.
On May 15, 2012, the team presented the inspection results to Mr. David Gronek 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:
25. Primary DC calculation(s) for safety-related buses.


=SUPPLEMENTAL INFORMATION=
26. PI&Ds for ECCS systems (electronic and 1/2 size - hard copy week of January 9).


==KEY POINTS OF CONTACT==
27. Index and Legend for PI&Ds.
===Licensee personnel===
:
: [[contact::C. Borelli]], Manager, WBN PRA
: [[contact::K. Dutton]], Director, Site Engineering
: [[contact::D. Gronek]], Plant Manager
: [[contact::D. Guinn]], Manager, Site Licensing
: [[contact::R. Kirkpatrick]], Manager, Design Engineering
: [[contact::T. Morgan]], Site Licensing
: [[contact::W. Nesmith]], Design Engineering


===NRC personnel===
28. Copy of design bases documents for ECCS systems.
: [[contact::R. Nease]], Chief, Engineering Branch Chief 1, Division of Reactor Safety, Region II
: [[contact::S. Schaeffer]], Chief, Project Branch 6, Division of Reactor Project, Region II
: [[contact::S. Sandal]], Senior Reactor Inspector, Division of Reactor Safety, Region II
: [[contact::B. Monk]], Senior Resident Inspector, Division of Reactor Projects, Watts Bar Resident Office
: [[contact::R. Mathew]], Team Leader, Division of Engineering, Office of Nuclear Reactor Regulation
: [[contact::G. Matharu]], Senior Electrical Engineer, Division of Engineering, Office of Nuclear Reactor
Regulation


==LIST OF ITEMS==
29. Copy of Operability determination procedure(s).
===OPENED, CLOSED AND DISCUSSED===
===Opened and Closed===
: 05000390/2012008-01 NCV Failure to Establish Test Procedures to Assure Satisfactory ACAS Performance during Design Basis Accidents [Section 1R21.2.4]
: 05000390/2012008-02 NCV Failure to Adequately Test the AFW Discharge Check Valves [Section 1R21.2.6]
: 05000390/2012008-03 NCV Inadequate Acceptance Criteria in Maintenance and Surveillance Procedures (5 Examples) [Section 1R21.2.9]


===Opened===
30. Copies of condition reports associated with findings from previous CDBI (if applicable).
: 05000390/2012008-04 URI Effect of System Harmonics on Degraded Voltage Relay Function [Section 1R21.2.9]


==LIST OF DOCUMENTS REVIEWED==
Enclosure 31. Index (procedure number, titles, and current revision) of station Emergency Operating Procedures (EOPs), Abnormal Operating Procedures (AOPs), and Annunciator Response Procedures (ARPs).


32. Contact information for a person to discuss PRA information prior to the information gathering trip (name, title, phone number, and e-mail address).
}}
}}

Latest revision as of 01:44, 13 January 2025

Notification of Watts Bar Nuclear Plant, Unit 1, Component Design Bases Inspection - NRC Inspection Report 05000390-12-008
ML112790114
Person / Time
Site: Watts Bar Tennessee Valley Authority icon.png
Issue date: 10/06/2011
From: Binoy Desai
NRC/RGN-II/DRS/EB1
To: Krich R
Tennessee Valley Authority
References
IR-12-008
Download: ML112790114 (7)
v • d • e


Text

October 06, 2011

SUBJECT:

NOTIFICATION OF WATTS BAR NUCLEAR PLANT, UNIT 1, COMPONENT DESIGN BASES INSPECTION - NRC INSPECTION REPORT 05000390/2012008

Dear Mr. Krich:

The purpose of this letter is to notify you that the U.S. Nuclear Regulatory Commission (NRC)

Region II staff will conduct a component design bases inspection at your Watts Bar Nuclear Plant during the weeks of January 30 - February 3, February 13 - 17, and February 27 - March 2, 2012. The inspection team will be led by Shane Sandal, a Senior Reactor Inspector from the NRC's Region II Office. This inspection will be conducted in accordance with the baseline inspection procedure, Procedure 71111.21, Component Design Bases Inspection, issued December 6, 2010.

The inspection will evaluate the capability of risk significant/low margin components to function as designed and to support proper system operation. The inspection will also include a review of selected operator actions, operating experience, and modifications.

From an email reply on September 29, 2011, Mr. Sandal confirmed with Mr. Riedl of your staff, arrangements for an information gathering site visit and the three-week onsite inspection. The schedule is as follows:

  • Information gathering visit: Week of January 9 - 13, 2012

Onsite weeks: January 30 - February 3, February 13 - 17, and February 27 -

March 2, 2012

The purpose of the information gathering visit is to meet with members of your staff to identify risk-significant components and operator actions. Information and documentation needed to support the inspection will also be identified. Mr. John Hanna, a Region II Senior Reactor Analyst, will accompany Mr. Sandal during the information gathering visit to review probabilistic risk assessment data and identify risk significant components which will be examined during the inspection.

The enclosure lists documents that will be needed prior to the information gathering visit.

Please provide the referenced information to the Region II office by January 2, 2012. Contact Mr. Sandal with any questions concerning the requested information. The inspectors will try to

TVA

minimize your administrative burden by specifically identifying only those documents required for inspection preparation.

Additional documents will be requested during the information gathering visit. The additional information will need to be made available to the team in the Region II office prior to the inspection teams preparation week of January 23, 2012. Mr. Sandal, will also discuss the following inspection support administrative details: availability of knowledgeable plant engineering and licensing personnel to serve as points of contact during the inspection; method of tracking inspector requests during the inspection; licensee computer access; working space; arrangements for site access; and other applicable information.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Thank you for your cooperation in this matter. If you have any questions regarding the information requested or the inspection, please contact Mr. Sandal at (404) 997-4513 or me at (404) 997-4519.

Sincerely,

/RA/

Binoy B. Desai, Chief

Engineering Branch 1

Division of Reactor Safety Docket Nos.: 50-390 License Nos.: NPF-90

Enclosure: Information Request for Watts Bar Nuclear Plant -

Component Design Bases Inspection

cc w/enc/: (See page 3)

___ML112790114__________________

X SUNSI REVIEW COMPLETE OFFICE RII:DRS RII:DRS

SIGNATURE

/RA/

/RA/

NAME S. SANDAL B. DESAI

DATE 10/ 03 /2011 10/ 06 /2011

E-MAIL COPY?

YES NO YES NO YES NO YES NO YES NO YES NO YES NO

TVA

cc w/encl:

D. E. Grissette Site Vice President Watts Bar Nuclear Plant Tennessee Valley Authority Electronic Mail Distribution

G. A. Boerschig Plant Manager Watts Bar Nuclear Plant, MOB 2R-WBN Tennessee Valley Authority Electronic Mail Distribution

C. J. Riedl Acting Manager, Licensing Watts Bar Nuclear Plant, ADM 1L-WBN Tennessee Valley Authority P.O. Box 2000 Spring City, TN 37381

J. W. Shea Manager, Corp. Nuclear Licensing - WBN Tennessee Valley Authority Electronic Mail Distribution

E. J. Vigluicci Assistant General Counsel Tennessee Valley Authority Electronic Mail Distribution

W. D. Crouch Licensing Manager, Unit 2 Watts Bar Nuclear Plant, EQB 1B-WBN Tennessee Valley Authority P.O. Box 2000 Spring City, TN 37381

County Mayor P.O. Box 156 Decatur, TN 37322

County Executive 375 Church Street Suite 215 Dayton, TN 37321

Tennessee Department of Environment &

Conservation Division of Radiological Health 401 Church Street Nashville, TN 37243

Senior Resident Inspector U.S. Nuclear Regulatory Commission Watts Bar Nuclear Plant U.S. Nuclear Regulatory Commission 1260 Nuclear Plant Road Spring City, TN 37381-2000

Ann Harris 341 Swing Loop Rockwood, TN 37854

Enclosure INFORMATION REQUEST FOR WATTS BAR NUCLEAR PLANT COMPONENT DESIGN BASES INSPECTION

Please provide the information electronically in.pdf files, Excel, or other searchable format on CDROM (or FTP site, Sharepoint, etc.) The CDROM (or website) should be indexed and hyperlinked to facilitate ease of use.

1.

From your most-recent Probabilistic Safety Analysis (PSA) excluding external events and fires:

a. Two risk rankings of components from your site-specific probabilistic safety analysis (PSA): one sorted by Risk Achievement Worth (RAW), and the other sorted by Birnbaum Importance

b. A list of the top 500 cutsets

2.

From your most-recent Probabilistic Safety Analysis (PSA) including external events and fires:

a. Two risk rankings of components from your site-specific Probabilistic Safety Analysis (PSA): one sorted by Risk Achievement Worth (RAW), and the other sorted by Birnbaum Importance

b. A list of the top 500 cutsets

3.

Risk ranking of operator actions from your site specific PSA sorted by RAW. Provide human reliability worksheets for these items.

4.

List of time critical operator actions with a brief description of each action.

5.

List of Emergency and Abnormal Operating Procedures revised (significant) since April 1, 2010 with a brief description of each revision.

6.

List of components with low design margins (i.e., pumps closest to the design limit for flow or pressure, diesel generator close to design required output, heat exchangers close to rated design heat removal, MOV risk-margin rankings, etc.) and associated evaluations or calculations.

7. List of station operating experience evaluations/reviews performed and documented in the stations corrective action program for industry events and safety related equipment failures/vulnerabilities [as communicated by NRC generic communications, industry communications, 10 CFR part 21 notifications, etc.] since April 1, 2010.

8.

List and brief description of safety related SSC design modifications implemented since April 1, 2010.

9.

List and brief description of common-cause component failures that have occurred since April 1, 2010.

Enclosure 10. List and brief description of operability evaluations completed since April 1, 2010.

11. List of equipment on the sites Station Equipment Reliability Issues List, including a description of the reason(s) why each component is on that list and summaries (if available)

of your plans to address the issue(s).

12. List and brief description of equipment currently in degraded or nonconforming status as described in RIS 05-020.

13. List and reason for equipment classified in maintenance rule (a)(1) status since April 1, 2010 to present.

14. Copies of System Descriptions (or the like design basis documents) for Safety-Related Systems.

15. Copy of UFSAR(s).

16. Copy of Technical Specification(s).

17. Copy of Technical Specifications Bases.

18. Copy of Technical Requirements Manual(s).

19. List and brief description of Root Cause Evaluations that have been performed since

April 1, 2010.

20. In-service Testing Program Procedure(s).

21. Corrective Action Program Procedure(s).

22. One line diagram of electrical plant (electronic and full size - hard copy week of January 9).

23. Index and legend for electrical plant one-line diagrams.

24. Primary AC calculation(s) for safety-related buses.

25. Primary DC calculation(s) for safety-related buses.

26. PI&Ds for ECCS systems (electronic and 1/2 size - hard copy week of January 9).

27. Index and Legend for PI&Ds.

28. Copy of design bases documents for ECCS systems.

29. Copy of Operability determination procedure(s).

30. Copies of condition reports associated with findings from previous CDBI (if applicable).

Enclosure 31. Index (procedure number, titles, and current revision) of station Emergency Operating Procedures (EOPs), Abnormal Operating Procedures (AOPs), and Annunciator Response Procedures (ARPs).

32. Contact information for a person to discuss PRA information prior to the information gathering trip (name, title, phone number, and e-mail address).

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