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| issue date = 08/01/2014
| issue date = 08/01/2014
| title = IR 05000327-14-003 & 05000328-14-003; on 4/1/2014 - 6/30/2014; Sequoyah Nuclear Plant, Units 1 and 2; In-Service Inspection; Radiological Hazard Assessment and Exposure Controls
| title = IR 05000327-14-003 & 05000328-14-003; on 4/1/2014 - 6/30/2014; Sequoyah Nuclear Plant, Units 1 and 2; In-Service Inspection; Radiological Hazard Assessment and Exposure Controls
| author name = Bartley J H
| author name = Bartley J
| author affiliation = NRC/RGN-II/DRP/RPB6
| author affiliation = NRC/RGN-II/DRP/RPB6
| addressee name = Shea J W
| addressee name = Shea J
| addressee affiliation = Tennessee Valley Authority
| addressee affiliation = Tennessee Valley Authority
| docket = 05000327, 05000328
| docket = 05000327, 05000328
Line 14: Line 14:
| page count = 49
| page count = 49
}}
}}
See also: [[followed by::IR 05000327/2014003]]
See also: [[see also::IR 05000327/2014003]]


=Text=
=Text=
{{#Wiki_filter:
{{#Wiki_filter:UNITED STATES
[[Issue date::August 1, 2014]]
                                NUCLEAR REGULATORY COMMISSION
                                                REGION II
                            245 PEACHTREE CENTER AVENUE NE, SUITE 1200
                                      ATLANTA, GEORGIA 30303-1257
                                          August 1, 2014
Mr. Joseph W. Shea
Vice President, Nuclear Licensing
Tennessee Valley Authority
1101 Market Street, LP 3D-C
Chattanooga, TN 37402-2801
SUBJECT: SEQUOYAH NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT
              05000327/2014003 AND 05000328/2014003
Dear Mr. Shea:
On June 30, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at
your Sequoyah Nuclear Plant, Units 1 and 2. On July 9, the NRC inspectors discussed the
results of this inspection with Mr. Simmons and other members of your staff. Inspectors
documented the results of this inspection in the enclosed inspection report.
NRC inspectors documented two findings which were determined to be of very low safety
significance (Green) in this report. These findings involved violations of NRC requirements.
If you contest the violation or significance of the NCV, you should provide a response within 30
days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear
Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with
copies to the Regional Administrator, Region II; the Director, Office of Enforcement, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector
at the Sequoyah Nuclear Plant.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a
response within 30 days of the date of this inspection report, with the basis for your
disagreement, to the Regional Administrator, Region II, and the NRC resident inspector at the
Sequoyah Nuclear Plant.


Mr. Joseph Vice President, Nuclear Licensing Tennessee Valley Authority 1101 Market Street, LP 3D-C Chattanooga, TN 37402-2801
J. Shea                                      2
In accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections,
Exemptions, Requests for Withholding, 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
NRCs Public Document Room or from the Publicly Available Records (PARS) component of
NRCs Agencywide Documents Access and Management System (ADAMS). ADAMS is
accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public
Electronic Reading Room).
                                              Sincerely,
                                                /RA/
                                              Jonathan H. Bartley, Chief
                                              Reactor Projects Branch 6
                                              Division of Reactor Projects
Docket Nos.: 50-327, 50-328
License Nos.: DPR-77, DPR-79
Enclosure: Inspection Report 050003272014003, 05000328/2014003
              w/Attachment: Supplementary Information
cc via ListServ distribution


SUBJECT: SEQUOYAH NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT 05000327/2014003 AND 05000328/2014003


==Dear Mr. Shea:==
_________________________                    SUNSI REVIEW COMPLETE        FORM 665 ATTACHED
On June 30, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Sequoyah Nuclear Plant, Units 1 and 2. On July 9, the NRC inspectors discussed the results of this inspection with Mr. Simmons and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report.
OFFICE            RII:DRP        RII:DRP      RII:DRS          RII:DRP        RII:DRS        RII:DRP        RII:DRSP
SIGNATURE          Via email      Via email    Via email        JHB /RA for/    Via email      Via email      Via email
NAME              GSmith        WDeschaine    PBraaten        CKontz          RHamilton      WPursley        RKellner
DATE                  7/31/2014      7/30/2014    7/29/2014        7/31/2014      7/29/2014      7/31/2014      7/30/20148/
E-MAIL COPY?        YES    NO    YES      NO  YES      NO      YES      NO    YES      NO    YES      NO    YES/2014 NO
OFFICE            RII:DRP        RII:DRS      RII:DRS          RII:DRP        RII:DRP
SIGNATURE          Via email      Via email    Via email        Via email      JHB /RA/
NAME              AButcavage    ASengupta    BCollins        JHamman        JBartley
DATE                  7/31/2014      7/29/2014    7/29/2014        7/31/2014      8/1/20148/
E-MAIL COPY?        YES    NO    YES      NO  YES      NO      YES      NO    YES/2014 NO    YES      NO    YES      NO
       
J. Shea                                  3
Letter to Joseph W. Shea from Jonathan H. Bartley dated August 1, 2014.
SUBJECT: SEQUOYAH NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT
              05000327/2014003 AND 05000328/2014003
Distribution:
D. Gamberoni, RII
L. Douglas, RII
OE Mail
RIDSNRRDIRS
PUBLIC
RidsNrrPMSequoyah Resource


NRC inspectors documented two findings which were determined to be of very low safety significance (Green) in this report. These findings involved violations of NRC requirements. If you contest the violation or significance of the NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, U.S.
              U. S. NUCLEAR REGULATORY COMMISSION
                                  REGION II
Docket Nos.:        50-327, 50-328
License Nos.:      DPR-77, DPR-79
Report Nos.:        05000327/2014003, 05000328/2014003
Licensee:          Tennessee Valley Authority (TVA)
Facility:          Sequoyah Nuclear Plant, Units 1 and 2
Location:          Sequoyah Access Road
                    Soddy-Daisy, TN 37379
Dates:              April 1 - June 30, 2014
Inspectors:        G .Smith, Senior Resident Inspector
                    W. Deschaine, Resident Inspector
                    P. Braaten, Reactor Inspector (1R04)
                    C. Kontz, Senior Project Engineer (1R05, 1R11, 1R18)
                    R. Hamilton, Senior Health Physicist (2RS02)
                    W. Pursley, Health Physicist (2RS01, 2RS03, 2RS04)
                    R. Kellner, Health Physicist (2RS05)
                    A. Butcavage, Reactor Inspector (1R08)
                    A. Sengupta, Reactor Inspector (1R08)
                    B. Collins, Reactor Inspector (4OA5)
Approved by:        Jonathan H. Bartley, Chief
                    Reactor Projects Branch 6
                    Division of Reactor Projects
                                                                        Enclosure


Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the Sequoyah Nuclear Plant.
                                            SUMMARY
IR 05000327/2014-003, 05000328/2014-003; 4/1-6/30/2014; Sequoyah Nuclear Plant, Units 1
and 2; In-Service Inspection; Radiological Hazard Assessment and Exposure Controls
The report covered a three-month period of inspection by resident and regional inspectors. Two
findings/violations were identified. The significance of most findings is indicated by their color
(Green, White, Yellow, Red) using 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," Revision 4, dated December 2006.
A.      NRC-Identified and Self-Revealing Findings
        Cornerstone: Mitigating Systems
        *  Green: An NRC-identified Green non-cited violation (NCV) of 10 CFR 50.55a(g)(4),
            Inservice Inspection Requirements was identified for the licensees failure to
            perform visual examinations of the control rod drive mechanism (CRDM), American
            Society of Mechanical Engineers (ASME) Class 1, seismic plate supports as required
            by the ASME Code, Section XI. The licensee entered this issue into their corrective
            action program (CAP) as Problem Evaluation Report (PER) 889400. The licensee
            developed an operability evaluation and concluded that the supports remained
            functional. The licensee also initiated corrective actions to perform the required
            visual examinations of the CRDM seismic plate supports before the end of the
            current inservice inspection (ISI) interval in April 2016.
            The finding was more than minor because it was associated with the protection
            against external factors attribute of the mitigating systems cornerstone, and affected
            the cornerstone objective to ensure availability, reliability, and capability of systems
            that respond to initiating events to prevent undesirable consequence. The inspectors
            screened this finding as Green because the finding did not involve the loss or
            degradation of equipment or function specifically designed to mitigate a seismic
            initiating event. A crosscutting aspect was not assigned to this finding in accordance
            with IMC 0612, Appendix B, because the exclusion of the CRDM seismic plate
            supports from the scope of the ISI Program occurred outside of the nominal 3-year
            period for present performance, and therefore it was not reflective of present
            licensee performance. (Section 1R08)
        Cornerstone: Occupational Radiation Safety
        *  Green: The inspectors identified a Green, self-revealing, NCV of Technical
            Specification (TS) 6.12.1, High Radiation Area, for two examples where workers
            made entries into High Radiation Areas (HRA) on May 16, 2014, without meeting the
            entry requirements specified therein. Specifically, these workers, while performing
            decontamination activities and moving materials in the upper reactor containment,
            entered a posted HRA: 1) without knowledge of the current radiological conditions in
                                                                                            Enclosure


If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region II, and the NRC resident inspector at the Sequoyah Nuclear Plant. In accordance with Title 10 of the Code of Federal Regulations 2.390, "Public Inspections, Exemptions, Requests for Withholding," 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 NRC's Public Document Room or from the Publicly Available Records (PARS) component of NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
                                          3
 
      the actual work area, 2) not using a radiological work permit (RWP) approved for
Sincerely,/RA/
      HRA entry, and 3) without wearing the prescribed electronic dosimetry for an HRA.
Jonathan H. Bartley, Chief Reactor Projects Branch 6 Division of Reactor Projects Docket Nos.: 50-327, 50-328 License Nos.: DPR-77, DPR-79
      The licensee entered these events into the Corrective Action Program (CAP) as
 
      Problem Evaluation Reports (PERs) Numbers 886668 and 886160. Immediate
===Enclosure:===
      corrective actions included restricting worker access to the Radiologically Controlled
Inspection Report 050003272014003, 05000328/2014003
      Area (RCA) and issuance of communications to the site and within the Radiation
 
      Protection organization to reinforce roles in RWP adherence and access control.
===w/Attachment:===
      This finding was more than minor because it is associated with the Occupational
Supplementary Information cc via ListServ distribution
      Radiation Safety Cornerstone attribute of Human Performance and adversely affects
 
      the cornerstone objective of ensuring adequate protection of worker health and
=SUMMARY=
      safety from exposure to radiation from radioactive material during routine civilian
IR 05000327/2014-003, 05000328/2014-003; 4/1-6/30/2014; Sequoyah Nuclear Plant, Units 1 and 2; In-Service Inspection; Radiological Hazard Assessment and Exposure Controls
      nuclear reactor operation. The finding was not related to As Low As Reasonably
 
      Achievable planning, nor did it involve an overexposure or substantial potential for
The report covered a three-month period of inspection by resident and regional inspectors. Two findings/violations were identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using 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," Revision 4, dated December 2006.
      overexposure and the ability to assess dose was not compromised. Therefore, the
 
      finding was determined to be of very low safety significance (Green). This finding
===A. NRC-Identified and Self-Revealing Findings===
      involved the cross-cutting aspect of Human Performance, Avoid Complacency [H.12]
 
      because workers failed to apply appropriate error reduction tools during participation
===Cornerstone: Mitigating Systems===
      in the pre-job brief and prior to crossing the HRA boundaries. (2RS1)
* Green:  An NRC-identified Green non-cited violation (NCV) of 10 CFR 50.55a(g)(4), "Inservice Inspection Requirements" was identified for the licensee's failure to perform visual examinations of the control rod drive mechanism (CRDM), American Society of Mechanical Engineers (ASME) Class 1, seismic plate supports as required by the ASME Code, Section XI. The licensee entered this issue into their corrective action program (CAP) as Problem Evaluation Report (PER) 889400. The licensee developed an operability evaluation and concluded that the supports remained functional. The licensee also initiated corrective actions to perform the required visual examinations of the CRDM seismic plate supports before the end of the current inservice inspection (ISI) interval in April 2016. The finding was more than minor because it was associated with the protection against external factors attribute of the mitigating systems cornerstone, and affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequence. The inspectors screened this finding as Green because the finding did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic initiating event. A crosscutting aspect was not assigned to this finding in accordance with IMC 0612, Appendix B, because the exclusion of the CRDM seismic plate supports from the scope of the ISI Program occurred outside of the nominal 3-year period for present performance, and therefore it was not reflective of present licensee performance.  (Section 1R08) 
B. Licensee-Identified Violations
 
  None.
===Cornerstone: Occupational Radiation Safety===
                                                                                    Enclosure
* Green: The inspectors identified a Green, self-revealing, NCV of Technical Specification (TS) 6.12.1, "High Radiation Area", for two examples where workers made entries into High Radiation Areas (HRA) on May 16, 2014, without meeting the entry requirements specified therein. Specifically, these workers, while performing decontamination activities and moving materials in the upper reactor containment, entered a posted HRA: 1) without knowledge of the current radiological conditions in the actual work area, 2) not using a radiological work permit (RWP) approved for HRA entry, and 3) without wearing the prescribed electronic dosimetry for an HRA. The licensee entered these events into the Corrective Action Program (CAP) as Problem Evaluation Reports (PERs) Numbers 886668 and 886160. Immediate corrective actions included restricting worker access to the Radiologically Controlled Area (RCA) and issuance of communications to the site and within the Radiation Protection organization to reinforce roles in RWP adherence and access control. This finding was more than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute of Human Performance and adversely affects the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The finding was not related to As Low As Reasonably Achievable planning, nor did it involve an overexposure or substantial potential for overexposure and the ability to assess dose was not compromised. Therefore, the finding was determined to be of very low safety significance (Green). This finding involved the cross-cutting aspect of Human Performance, Avoid Complacency [H.12] because workers failed to apply appropriate error reduction tools during participation in the pre-job brief and prior to crossing the HRA boundaries. (2RS1)  
 
===B. Licensee-Identified Violations===
None.
 
=REPORT DETAILS=


                                        REPORT DETAILS
Summary of Plant Status:
Summary of Plant Status:
Unit 1 operated at or near 100 percent rated thermal power (RTP) for the entire inspection period.
Unit 1 operated at or near 100 percent rated thermal power (RTP) for the entire inspection
 
period.
Unit 2 operated at or near 100 percent RTP until April 12, 2014, when the unit entered a power coast down period. On May 12, with the unit at 76 percent RTP, Unit 2 was shut down for a refueling outage. Unit 2 returned to 100 percent RTP on June 21, where it operated for the remainder of the inspection period.
Unit 2 operated at or near 100 percent RTP until April 12, 2014, when the unit entered a power
 
coast down period. On May 12, with the unit at 76 percent RTP, Unit 2 was shut down for a
==REACTOR SAFETY==
refueling outage. Unit 2 returned to 100 percent RTP on June 21, where it operated for the
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity 1R04 Equipment Alignment
remainder of the inspection period.
 
1.      REACTOR SAFETY
===.1 Partial System Walkdown===
        Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
 
1R04 Equipment Alignment
====a. Inspection Scope====
.1     Partial System Walkdown
The inspectors performed partial walkdowns of the following two systems to verify the operability of redundant or diverse trains and components when safety equipment was inoperable. The inspectors focused on identification of discrepancies that could impact the function of the system and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, walked down control system components; and determined whether selected breakers, valves, and support equipment were in the correct position to support system operation. The inspectors also verified that the licensee had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the corrective action program (CAP). Documents reviewed are listed in the Attachment. This activity constituted two inspection samples.
  a.   Inspection Scope
* Spent fuel pool cooling system during Unit 2 core empty period
        The inspectors performed partial walkdowns of the following two systems to verify the
* Unit 1 B-train High Head Safety Injection system during A-train planned maintenance
        operability of redundant or diverse trains and components when safety equipment was
 
        inoperable. The inspectors focused on identification of discrepancies that could impact
====b. Findings====
        the function of the system and, therefore, potentially increase risk. The inspectors
No findings were identified.
        reviewed applicable operating procedures, walked down control system components;
 
        and determined whether selected breakers, valves, and support equipment were in the
===.2 Complete System Walkdown===
        correct position to support system operation. The inspectors also verified that the
 
        licensee had properly identified and resolved equipment alignment problems that could
====a. Inspection Scope====
        cause initiating events or impact the capability of mitigating systems or barriers and
The inspectors performed a complete system walk down of the Unit 2 Main Steam and support systems to verify proper equipment alignment, to identify any discrepancies that could impact the function of the system and increase risk, and to verify that the licensee properly identified and resolved equipment alignment problems that could cause events or impact the functional capability of the system.
        entered them into the corrective action program (CAP). Documents reviewed are listed
 
        in the Attachment. This activity constituted two inspection samples.
The inspectors reviewed the Updated Final Safety Analysis Report (UFSAR), system procedures, system drawings, and system design documents to determine the correct lineup and then examined system components and their configuration to identify any discrepancies between the existing system equipment lineup and the correct lineup. During the walkdown, the inspectors reviewed the following:
        *   Spent fuel pool cooling system during Unit 2 core empty period
* Valves were correctly positioned and did not exhibit leakage that would impact the functions of any given valve.
        *   Unit 1 B-train High Head Safety Injection system during A-train planned maintenance
* Electrical power was available as required.
  b.  Findings
* Major system components were correctly labeled, lubricated, cooled, ventilated, etc.
        No findings were identified.
* Hangers and supports were correctly installed and functional.
                                                                                          Enclosure
* Essential support systems were operational.
* Ancillary equipment or debris did not interfere with system performance.
* Valves were locked as required by the locked valve program.
* Major system components were correctly labeled.
* Visible cabling appeared to be in good material condition. In addition, the inspectors reviewed outstanding maintenance work requests and design issues on the system to determine whether any condition described in those work requests could adversely impact current system operability. Documents reviewed are listed in the Attachment. This activity constituted one inspection sample.
 
====b. Findings====
No findings were identified.
{{a|1R05}}
==1R05 Fire Protection==
 
===.1 Fire Protection Tours===
 
====a. Inspection Scope====
The inspectors conducted a tour of the five areas important to safety listed below to assess the material condition and operational status of fire protection features. The inspectors evaluated whether: combustibles and ignition sources were controlled in accordance with the licensee's administrative procedures; fire detection and suppression equipment was available for use; passive fire barriers were maintained in good material condition; and compensatory measures for out-of-service, degraded, or inoperable fire protection equipment were implemented in accordance with the licensee's fire plan. Documents reviewed are listed in the Attachment. This activity constituted five inspection samples.
* Control Building Elevation 669 (Mechanical Equipment Room, 250 VDC Battery and Battery Board Rooms)
* Control Building Elevation 685 (Auxiliary Instrument Rooms)
* Turbine Building Elevation 706
* Control Building Elevation 706 (Cable Spreading Room)
* Control Building Elevation 732 (Mechanical Equipment Room and Relay Room)
 
====b. Findings====
No findings were identified.
{{a|1R06}}
==1R06 Flood Protection Measures==
Annual Review of Cables Located in Underground Bunkers/Manholes
 
====a. Inspection Scope====
The inspectors conducted a review of licensee inspections of safety-related cables located in underground bunkers/manholes subject to flooding. Specifically, inspectors reviewed maintenance records of inspections for the previous 12 months to determine if water was present and, if found, whether it would affect safety-related system operation.
 
In addition, the inspectors reviewed the licensee's corrective action program (CAP) to ensure that the licensee was identifying underground cabling issues and that they were properly addressed for resolution. Documents reviewed are listed in the Attachment.
 
This activity constituted one inspection sample.
 
====b. Findings====
No findings were identified.
{{a|1R08}}
==1R08 Inservice Inspection Activities==
 
====a. Inspection Scope====
Non-Destructive Examination Activities and Welding Activities:  From May 19, 2014, through May 30, 2014, the inspectors conducted an onsite review of the implementation of the licensee's in-service inspection (ISI) program for monitoring degradation of the reactor coolant system (RCS), risk-significant piping and components, and containment systems in Unit 2. The inspectors' activities included a review of selected samples of non-destructive examinations (NDE) to evaluate compliance with the applicable edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), Section XI, and to verify that indications and defects (if present) were appropriately evaluated and dispositioned, in accordance with the requirements of the ASME Section XI acceptance standards. The inspectors directly observed the following NDE, mandated by the ASME Code, to evaluate compliance with the ASME Code Section XI, and Section V requirements, and if any indications or defects were detected, to evaluate if they were dispositioned in accordance with the ASME Code, or an NRC-approved alternative requirement.
* Visual Examination (VT) - 3,  Steam Generator (SG) Upper Lateral Restraint  SGH-4-1, ASME Code Class 2
* General Visual Examination, Containment Moisture Barrier, Examination Category  E-A, Item No E1.30  Inspectors directly observed the calibration of Ultrasonic Test (UT) equipment, and later reviewed UT examination results for welds associated with a feedwater elbow attachment to the SG safe end.
* Equipment Calibration for UT, Feedwater Piping Weld FDF-011A, ASME Code  Class 2, Augmented Inspection
* Equipment Calibration for UT, Feedwater Piping Weld FDF-010C, ASME Code  Class 2, Augmented Inspection The inspectors reviewed records of the following NDE inspections and methods mandated by the ASME Code Section XI or augmented inspections, in order to evaluate compliance with the ASME Code Section XI and Section V requirements, and if any indications and defects were detected, to evaluate if they were dispositioned in accordance with the ASME Code or NRC-approved alternative requirements.
* VT-3, Pipe Support, 2-CVCH-585, ASME Code Class 2
* VT-3, Pipe Support, 2-CVCH-584, ASME Code Class 2
* VT-3, Pipe Support, 2-CVCH-586, ASME Code Class 2
* Penetrant Testing (PT), Integral Attachment Weld, 2-CVCH-006-IA, Code Class 1
* UT, Feedwater Piping Weld FDF-011A, ASME Code Class 2, (Augmented)
* UT, Feedwater Piping Weld FDF-010C, ASME Code Class 2, (Augmented)  The inspectors reviewed the following surface examination records with recordable indications that were analytically evaluated and accepted for continued service, against the ASME Code Section XI, or an NRC-approved alternative.
* PT, Lug to Pipe Weld, 2-SIH-020-IA, ASME Code Class 1 No ASME Class 1, 2, or 3 welding activities were in progress during the NRC ISI inspector site visit. Therefore, the inspectors reviewed the previously completed welding activity work order (WO), referenced below, in order to evaluate compliance with the intent of procedures, and the ASME Code. Specifically, the inspectors reviewed the WO package, the WO VT-2 leakage examination requirements and results.
* WO No. 112354373, SQN-2-VLV-0012-0817, Valve Replacement, ASME Class 2  Pressurized Water Reactor Vessel Upper Head Penetration Inspection Activities:  For the Unit 2 reactor vessel head, a full bare metal visual (BMV) examination was not required this outage pursuant to 10 CFR 50.55a. Therefore, no reviews were conducted for this inspection attribute. A volumetric examination of the Unit 2 vessel upper head penetration (VUHP) was required this outage. Therefore, inspectors observed and reviewed a sample of the Unit 2 UT examination results, which included NDE reports for VUHP Nos. 53, 56, and 60. The inspectors also performed a comparison of the current UT results to the previous UT examination results for the sample penetrations. These comparisons were used to determine if the activities, including the disposition of indications and defects, were conducted in accordance with the requirements of ASME Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D). In particular, the inspectors evaluated if the required UT examination scope/coverage was achieved and limitations (if applicable) were recorded in accordance with the licensee procedures. The licensee did not identify any relevant indications that needed to be accepted by analysis for continued service during the UT examination. Additionally, the licensee did not perform any welding repairs to the vessel head penetrations since the beginning of the last Unit 2 refueling outage; therefore, no NRC review was completed for these inspection procedure attributes.
 
Boric Acid Corrosion Control Inspection Activities:  The inspectors reviewed the licensee's boric acid corrosion control (BACC) program activities, to ensure implementation with commitments made in response to NRC Generic Letter 88-05, "Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary," and applicable industry guidance documents. Specifically, the inspectors performed an onsite record review of procedures and the results of the licensee's containment walkdown inspections performed during the current spring refueling outage. The inspectors also interviewed the BACC program owner, conducted an independent walkdown of two levels of containment, in order to evaluate compliance with the licensee's BACC program requirements, and verify that degraded or non-conforming conditions, such as boric acid leaks, were properly identified and corrected in accordance with the licensee's BACC and corrective action program (CAP).
 
The inspectors reviewed the following problem evaluation report (PER), and associated corrective actions related to evidence of boric acid leakage, to evaluate if the corrective actions completed were consistent with the requirements of the ASME Code Section XI, and 10 CFR Part 50, Appendix B, Criterion XVI,  and Industry Guidelines.
* PER 743110, Degraded Non-conforming condition for PDO on RCS leakage and Ice on Intermediate Deck Doors (IDD), 7/19/13 The inspectors reviewed the following engineering evaluations completed for evidence of boric acid leakage to determine if degraded components were documented in the CAP. The inspectors also reviewed corrective actions for any degraded components to determine if they met the applicable requirements of the ASME Code, Section XI, and/or NRC-approved alternatives.
* PER 888330, Boric Acid Leakage Evaluation, Reactor Cavity Nozzle Cover Seal Leaking into Keyway, 6/24/14
* PER 890230, Evaluation of Boric Acid Corrosion Damage, 2-SNUB-068-RCH30, 6/7/14
* SR 889942, Determine Available Margins in Pipe Support Attributes, 2-RCH-0028, 5/26/14  Steam Generator Tube Inspection Activities:  The inspectors reviewed the eddy current (EC) examination activities performed in Unit 2 SGs 1, 2, 3, and 4 during the end-of-cycle 19 refueling outage, to verify compliance with the licensee's Technical Specifications, ASME BPVC Section XI, and Nuclear Energy Institute (NEI) 97-06, "Steam Generator Program Guidelines."  The inspectors interviewed licensee personnel and vendor staff responsible for the SG inspection project, and reviewed documentation associated with the SG inspections and integrity assessments, as described in this report section.
 
The inspectors reviewed the scope of the EC examinations to verify that known and potential areas of tube degradation were inspected. The inspectors also verified that inspection scope expansion criteria were implemented based on inspection results, as directed by the Electric Power Research Institute (EPRI) "Pressurized Water Reactor Steam Generator Examination Guidelines, Revision 7."
 
The inspectors reviewed documentation for a sample of EC data analysts, EC probes, and EC testers to verify that personnel and equipment were qualified to detect the existing and potential degradation mechanisms applicable to Sequoyah's SG tubes, in accordance with the EPRI Examination Guidelines. This review included a sample of site-specific Examination Technique Specification Sheets (ETSSs) that were selected based on plant-specific and industry operating experience, to ensure that their qualification and site-specific implementation were consistent with Appendix H or I of the EPRI Examination Guidelines. The selected ETSSs for review consisted of bobbin and rotating probe techniques that were used to detect wear at the tube interface with support structures (i.e., tube support plates, anti-vibration bar (AVB), and flow distribution baffle plate), and wear associated with foreign objects.
 
The inspectors also reviewed a sample of EC data with a qualified data analyst to confirm that data analysis was performed in accordance with the applicable ETSS's and site-specific analysis guidelines. The inspectors verified that the equipment configuration was consistent with the essential parameters of the applicable technique.
 
The inspectors also verified that recordable indications were detected and sized in accordance with vendor procedures. As part of the EC data review, the inspectors verified that the EC indications on each selected tube were consistent with historical data relative to the number of indications, location, and size. The sample of EC data selected for review is listed below:
Steam Generator Tube Row/Column Eddy Current Probe Indication Type 2 R93/C59 Bobbin AVB wear 2 R93/C59 MRPC + point AVB Wear 2  R93/C59 Array  AVB Wear  2 R89/C59  Bobbin  AVB Wear  2 R89/C59  MRPC + point  AVB Wear 4 R93/C47 Bobbin Proximity Signal 4 R93/C47 Array Proximity Signal 2 R5/C101 Bobbin Distorted Support Signal 2 R5/C101 MRPC + point Distorted Support Signal  The inspectors selected a sample of wear degradation mechanisms from the Steam Generator Degradation Assessment, and verified that the in-situ pressure testing criteria were determined, in accordance with the EPRI Tube Integrity Guidelines. Additionally, the inspectors reviewed EC indication reports to determine whether tubes with relevant indications were appropriately screened for in-situ pressure testing.
 
The inspectors compared the recent EC examination results with the last Operational Assessment report for SGs to assess the licensee's prediction capability for maximum tube degradation, and number of tubes with indications. The inspectors verified that the licensee's evaluation was conservative and that current examination results were bound by the Operational Assessment projections.
 
The inspectors also compared past examination results discussed in the latest Degradation Assessment with the recent EC examination results to verify that new degradation mechanisms, if any, were identified and evaluated before plant startup. The review of EC examination results included the disposition of potential loose part indications on the SG secondary side, to verify that corrective actions for evaluating and retrieving loose parts were consistent with the EPRI Guidelines. The inspectors also reviewed a sample of primary-to-secondary leakage data for Unit 2 to confirm that operational leakage in all SGs remained below the action level threshold during the previous operating cycle.
 
Based on the review of the final EC examination results for all SGs and interviews with the licensee, the inspectors confirmed that no EC scope expansion was required, and none of the SG tubes examined met the criteria for plugging or in-situ pressure testing.
 
Furthermore, the inspectors interviewed licensee staff and reviewed a sample of secondary side visual inspection results for the SGs 1, 2, 3, and 4 upper bowl areas, to verify that potential areas of degradation based on site-specific operating experience were inspected, and appropriate corrective actions were taken to address degradation indications. This review included the results of Foreign Object Search and Retrieval (FOSAR) activities in all SGs, and an evaluation for loose parts in the secondary side of SGs 1, 2, 3, and 4.
 
Identification and Resolution of Problems:  The inspectors reviewed a sample of ISI-related problems which were identified by the licensee, and entered into the CAP as PERs. The inspectors reviewed the PERs to confirm that the licensee had appropriately described the scope of the problem, and had initiated corrective actions. The review also included the licensee's consideration and assessment of operating experience events applicable to the plant. The inspectors performed this review to ensure compliance with 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action" requirements. Documents reviewed are listed in the Attachment.
 
====b. Findings====
 
=====Introduction:=====
An NRC-identified Green NCV of 10 CFR 50.55a (g)(4), "Inservice Inspection Requirements" was identified for the licensee's failure to perform visual examinations of the control rod drive mechanism (CRDM), ASME Class 1, seismic plate supports as required by the ASME Code, Section XI.
 
=====Description:=====
The Sequoyah Unit 2 ISI program for the current interval (third interval) was prepared in accordance with the 2001 Edition of the ASME Section XI Code, with addenda through 2003. Article IWF-2000 of ASME Section XI, Table IWF-2500-1, Examination Category Item Number F1.40, requires, in part, periodic VT-3 visual examinations of supports (other than piping supports) in Class 1 components. The examinations provide reasonable assurance that the supports can continue to perform their intended function.
 
The CRDM assemblies are ASME Class 1 pressure retaining components that contain a series of seismic plate supports to ensure that the allowable design stress limits for the CRDM assemblies are not exceeded during a seismic event, which in turn provides reasonable assurance that the RCS pressure boundary and control rod function is maintained.
 
The inspectors identified that the Sequoyah Unit 2 ISI program did not meet the requirements of ASME Section XI in that the Class 1 CRDM seismic plate supports, and associated load path components, which meet the examination category F1.40, were not included in the scope of the program for the first, second, and third ISI intervals. The inspectors also identified that this issue applied to the Unit 1 ISI Program.
 
The licensee entered this issue into their CAP as PER 889400. The licensee developed an operability evaluation and concluded that the supports were operable but non-conforming. The evaluation considered previous dimensional verifications of the reactor vessel head lift rig components in the area of the CRDM seismic support plates, and as-found settings of the seismic plates from a modification project WO package associated with Unit 1 and 2 cables in the seismic plate area of the lift rig. The WO package included requirements to insert a gap gauge at each seismic plate screw pad gap to verify the correct gap was present on Unit 1. The results of Unit 1 as-found gap settings provided reasonable assurance that the as-found gap settings were adequate for Unit 2 based on the similarities in design, operating conditions, and implementation of outage maintenance activities. The evaluation also considered that no degradation of the lift rig intervening steel components in the support load path between the seismic plates and lift rig struts had been reported in previous outages through the CAP. The licensee also initiated corrective actions to perform the required visual examinations of the CRDM seismic plate supports before the end of the current ISI interval in April 2016.
 
=====Analysis:=====
Failure to perform the required visual examinations of the CRDM seismic plates and associated load path components, as required by the ASME Section XI Code, was a performance deficiency (PD). In accordance with Inspection Manual Chapter (IMC) 0612 Appendix B, "Issue Screening," the PD was more than minor because it was associated with the protection against external factors attribute of the mitigating systems cornerstone, and affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequence. Specifically, the licensee failed to perform examinations required to provide reasonable assurance that the support components can perform their intended function during design basis seismic events, and therefore maintain the stresses in the CRDM assembly within the allowable design limits, which in turn provides reasonable assurance that the RCS pressure boundary and control rod function is maintained. The inspectors screened this finding as Green in accordance with IMC 0609, Appendix A, Exhibit 2, Mitigating Systems, because the finding did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic initiating event. A crosscutting aspect was not assigned to this finding in accordance with IMC 0612, Appendix B, because the exclusion of the CRDM seismic plate supports from the scope of the ISI Program occurred outside of the nominal 3-year period for present performance, and therefore it was not reflective of present licensee performance.
 
=====Enforcement:=====
Title 10 CFR 50.55a(g)(4), "Inservice Inspection Requirements," requires in part that throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) that are classified as ASME Code Class 1, must meet the requirements, except design and access provisions, and preservice examination requirements set forth in Section XI of editions and addenda of the ASME BPVC that become effective subsequent to editions specified in paragraphs (g)(2) and (g)(3) of this Section, and that are incorporated by reference in paragraph (b) of this Section, to the extent practical within the limitations of design, geometry, and materials of construction of the components. Section XI of the ASME BPVC, 2001 Edition with 2003 Addenda, Table IWF-2500-1, Examination Category F-A Supports, requires a VT-3 examination of 100 percent of the ASME Class 1 supports, other than piping supports, every ISI Interval (examination item F1.40), as modified by Notes 1, 2, 3 and 5 of Table IWF-2500-1.
 
Contrary to the above, from initial commercial operation until present, the licensee failed to perform the required VT-3 examination of ASME Class 1 supports, other than piping supports, (i.e., seismic support plates and associated load path components) on the CRDM assemblies of Units 1 and 2. The licensee entered the issue into the CAP as PER 889400. The licensee initiated corrective actions to perform the required VT-3 examinations during the next refueling outage in order to restore compliance with   the 10 CFR 50.55a regulations. Because this violation was determined to be of very low safety significance (i.e., Green), and the licensee entered the issue in the CAP, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy, dated July 9, 2013. This finding will be tracked as NCV 05000327, 328/2014003-01, Failure to Perform Visual Examination of the Unit 1 and Unit 2 CRDM Seismic Plate Supports.
{{a|1R11}}
==1R11 Licensed Operator Requalification Program==
 
Quarterly Review
 
====a. Inspection Scope====
On June 24, 2014, the inspector observed an evaluated simulator scenario and the evaluators' critique of crew performance. The exercise was performed to provide practice to the operating crews in longer duration "CPE style" scenarios. The inspector observed crew performance in terms of:  communications; ability to take timely and proper actions; prioritizing, interpreting and verifying alarms; correct use and implementation of procedures, including the alarm response procedures; timely control board operation and manipulation, including high risk operator actions; oversight and direction provided by shift manager, including the ability to identify and implement appropriate Technical Specification (TS) action; and, group dynamics involved in crew performance. The inspector observed the ability of the licensee to administer the evaluation and quality of the evaluators' critique. The inspector observed scenario operations for simulator fidelity to verify that it matched actual plant response. Based on crew performance and scenario administration issues, the inspector also reviewed the follow-up actions taken to address operator deficiencies and identified administration issues. Documents reviewed are listed in the Attachment. This activity constituted one inspection sample.
 
====b. Findings====
No findings were identified
 
===.2 Quarterly Review of Licensed Operator Performance===
 
====a. Inspection Scope====
The inspectors observed and assessed licensed operator performance in the main control room during periods of heightened activity or risk. The inspectors reviewed various licensee policies and procedures such as OPDP-1, Conduct of Operations, NPG-SPP-10.0, Plant Operations, and 0-GO-5, Normal Power Operation. The inspectors utilized activities such as post-maintenance testing, surveillance testing, unplanned transients, infrequent plant evolutions, plant startups and shutdowns, reactor power and turbine load changes, and refueling and other outage activities to focus on the following conduct of operations as appropriate:
* operator compliance and use of procedures
* control board manipulations
* communication between crew members
* use and interpretation of plant instruments, indications and alarms
* use of human error prevention techniques
* documentation of activities, including initials and sign-offs in procedures
* supervision of activities, including risk and reactivity management
* pre-job briefs  Specifically, the inspectors observed licensed operator performance during the following activities:
* Unit 2 reactor shut down and cool down
* Unit 2 reactor start up  Documents reviewed are listed in the Attachment. This activity constituted one inspection sample.
 
====b. Findings====
No findings were identified  1R12 Maintenance Effectiveness
 
====a. Inspection Scope====
The inspectors reviewed the maintenance activities, issues, and/or systems listed below to verify the effectiveness of the licensee's activities in terms of: appropriate work practices; identifying and addressing common cause failures; scoping in accordance with 10 CFR 50.65(b); characterizing reliability issues for performance; trending key parameters for condition monitoring; charging unavailability for performance; classification in accordance with 10 CFR 50.65(a)(1) or (a)(2); appropriateness of performance criteria for structure, system, or components (SSCs) and functions classified as (a)(2); and appropriateness of goals and corrective actions for SSCs and functions classified as (a)(1). Documents reviewed are listed in the Attachment. This activity constituted one inspection sample.
* Cause Determination Evaluation 2741 associated with failure of flow switch (FS) 2-FS-74
 
====b. Findings====
No findings were identified.
{{a|1R13}}
==1R13 Maintenance Risk Assessments and Emergent Work Control==
 
====a. Inspection Scope====
The inspectors reviewed the following activities to determine whether appropriate risk assessments were performed prior to removing equipment from service for maintenance. The inspectors evaluated whether risk assessments were performed as required by 10 CFR 50.65(a)(4), and were accurate and complete. When emergent work was performed, the inspectors reviewed whether plant risk was promptly reassessed and managed. The inspectors also assessed whether the licensee's risk assessment tool use and risk categories were in accordance with Standard Programs and Processes Procedure NPG-SPP-07.1, "On-Line Work Management," Revision 3, and Instruction 0-TI-DSM-000-007.1, "Risk Assessment Guidelines," Revision 9. Documents reviewed are listed in the Attachment. The inspectors completed five samples.
* Unit 1 Yellow probabilistic safety assessment (PSA) risk associated with 1B Residual Heat Removal (RHR) pump planned maintenance
* emergent work due to failure of Individual Rod Position Indication (IRPI) E-5
* maintenance risk review U2R19 Outage Schedule
* emergent work due to failure of "CS" Component Cooling System (CCS) swing pump hand switch
* emergent work due to failure of Unit 2 vacuum breaker (2-30-573)
 
====b. Findings====
No findings were identified.
{{a|1R15}}
==1R15 Operability Evaluations==
 
====a. Inspection Scope====
For the four operability evaluations described in the PERs listed below, the inspectors evaluated the technical adequacy of the evaluations to ensure that TS operability was properly justified and the subject component or system remained available, such that no unrecognized increase in risk occurred. The inspectors compared the operability evaluations to UFSAR descriptions to determine if the system or component's intended function(s) were adversely impacted. In addition, the inspectors reviewed compensatory measures implemented to determine whether the compensatory measures worked as stated and the measures were adequately controlled. The inspectors also reviewed a sampling of PERs to assess whether the licensee was identifying and correcting any deficiencies associated with operability evaluations. Documents reviewed are listed in the Attachment. This activity constituted four inspection samples.
* PER 855557/864224:  Operation Decision Making Instruction (ODMI) for Unit 2 Power Range Overpower Rod Stop Deviation Alarms
* PER 886167:  ODMI for Unit 1 Cavity Seal Leakage
* PER 855850:  Past operability evaluation (POE) associated with 2B RHR 2-FS-74-24A failure
* PER 897994:  Prompt Determination of Operability (PDO) for Unit 2 Turbine Driven AFW pump
 
====b. Findings====
No findings were identified.
{{a|1R19}}
==1R19 Post-Maintenance Testing==
 
====a. Inspection Scope====
The inspectors reviewed the post-maintenance tests associated with the six work orders (WOs) listed below to assess whether procedures and test activities ensured system operability and functional capability. The inspectors reviewed the licensee's test procedure to evaluate whether:  the procedure adequately tested the safety function(s) that may have been affected by the maintenance activity, the acceptance criteria in the procedure were consistent with information in the applicable licensing basis and/or design basis documents, and the procedure had been properly reviewed and approved. The inspectors also witnessed the test or reviewed the test data to determine whether test results adequately demonstrated restoration of the affected safety function(s). Documents reviewed are listed in the Attachment. This activity constituted six inspection samples.
* WO 115149300, Rx Vessel Wide Range Level Failed High
* WO 115806034, Unit 1 Electric Pulse Repair of IRPI Connectors
* WO 114973816, Unit 1 RHR Mini Flow Valve environmental qualification maintenance and Inspection
* WO 113877775, RHR Return Valve Leak Rate Test for FCV-74-1 and FCV-74-2
* WO 113880726, SIS/RHR Hot Leg Check Valve Backseat Test
* WO 113875488, Post Maintenance Local Leak Rate Test (as-left) for 2-FCV-63-71, 2-FCV-63-84, & 2-FCV-63-23
 
====b. Findings====
No findings were identified.
{{a|1R20}}
==1R20 Refueling and Outage Activities==
Unit 2 Refueling Outage Cycle 19
 
====a. Inspection Scope====
For the Unit 2 refueling outage that began on May 12, the inspectors evaluated licensee activities in order to verify that the licensee considered risk in developing outage schedules, followed risk reduction methods developed to control plant configuration, developed mitigation strategies for the loss of key safety functions, and adhered to operating license and TS requirements that ensure defense-in-depth. The inspectors also walked down portions of Unit 2 not normally accessible during at-power operations to verify that safety-related and risk-significant SSCs were maintained in an operable condition. Specifically, between May 12 and June 30, the inspectors performed inspections and reviews of the following outage activities. This activity constituted one inspection sample for Refueling Activities.
* Outage Plan. The inspectors reviewed the outage safety plan and contingency plans to confirm that the licensee had appropriately considered risk, industry experience, and previous site-specific problems in developing and implementing a plan that assured maintenance of defense-in-depth.
* Reactor Shutdown. The inspectors observed the shutdown in the control room from the time the reactor was tripped until operators placed it on the RHR system for decay heat removal to verify that TS cool down restrictions were followed. The inspectors also toured the lower containment as soon as practicable after reactor shutdown to observe the general condition of the reactor coolant system (RCS), emergency core cooling system components, and to look for indications of previously unidentified leakage inside the polar crane wall.
* Licensee Control of Outage Activities. On a daily basis, the inspectors attended the licensee outage turnover meeting, reviewed PERs, and reviewed the defense-in-depth status sheets to verify that status control was commensurate with the outage safety plan and in compliance with the applicable TS when taking equipment out of service. The inspectors further toured the main control room and areas of the plant daily to ensure that the following key safety functions were maintained in accordance with the outage safety plan and TS:  electrical power, decay heat removal, spent fuel cooling, inventory control, reactivity control, and containment closure. The inspectors also observed a tag-out (2-TO-2014-0039, Tag-out of 2B-B Centrifugal Charging Pump) to verify that the equipment was appropriately configured to safely support the work and testing. To ensure that RCS level instrumentation was properly installed and configured to give accurate information, the inspectors reviewed the installation of the Mansell level monitoring system. Specifically, the inspectors discussed the system with engineering, walked it down to verify that it was installed in accordance with procedures and adequately protected from inadvertent damage, verified that Mansell indication properly overlapped with pressurizer level instruments during pressurizer drain-down, verified that operators properly set level alarms to procedurally required set-points, and verified that the system consistently tracked RCS level while lowering to reduced inventory conditions. The inspectors also observed operators compare the Mansell indications with locally-installed ultrasonic level indicators during entry into reduced inventory conditions.
* Refueling Activities. The inspectors observed fuel movement at the spent fuel pool and at the refueling cavity in order to verify compliance with TS and that each assembly was properly tracked from core offload to core reload. In order to verify proper licensee control of foreign material, the inspectors verified that personnel were properly checked before entering any foreign material exclusion (FME) areas, reviewed FME procedures, and verified that the licensee followed the procedures. To ensure that fuel assemblies were loaded in the core locations specified by the design, the inspectors independently reviewed the recording of the licensee's final core verification.
* Reduced Inventory and Mid-Loop Conditions. Prior to the outage, the inspectors reviewed the licensee's commitments to Generic Letter 88-17. Before entering reduced inventory conditions the inspectors verified that these commitments were in place, that plant configuration was in accordance with those commitments, and that distractions from unexpected conditions or emergent work did not affect operator ability to maintain the required reactor vessel level. The inspectors verified that licensee procedures for closing the containment upon a loss of decay heat removal were in effect, that operators were aware of how to implement the procedures, and that other personnel were available to close containment penetrations, if needed. In order to reduce outage risk, the licensee elected to not put the plant into mid-loop conditions during this particular refueling outage.
* Heatup and Startup Activities. The inspectors toured the containment prior to reactor startup to verify that debris that could affect the performance of the containment sump had not been left in the containment. The inspectors reviewed the licensee's mode-change checklists to verify that appropriate prerequisites were met prior to changing TS modes. Prior to plant startup, the inspectors performed a detailed tour of containment to ensure no debris existed that could affect containment sump performance given a design basis accident. The inspectors also inspected the primary system in containment during Mode 3 with the plant at normal operating pressure and temperature in order to verify the leak tightness of the RCS.
 
====b. Findings====
No findings were identified.
{{a|1R22}}
==1R22 Surveillance Testing==
 
====a. Inspection Scope====
For the 11 surveillance tests identified below, the inspectors assessed whether the SSCs involved in these tests satisfied the requirements described in the TS surveillance requirements, the UFSAR, applicable licensee procedures, and whether the tests demonstrated that the SSCs were capable of performing their intended safety functions.
 
This was accomplished by witnessing testing and/or reviewing the test data. Documents reviewed are listed in the Attachment. This activity constituted 11 inspection samples.
 
In-Service Tests:
* 2-SI-SXP-063-202.0, Safety Injection Pumps 2A-A and 2B-B Comprehensive Performance and Check Valve Test, Revision 5 Routine Surveillance Tests:
* 0-SI-OPS-092-078.0 Power Range Neutron Flux Channel Calibration by Heat Balance Comparison, Revision 23
* 0-SI-NUC-000-038.0 Unit 2 Shutdown Margin, Revision 75
* 2-SI-OPS-082-026.B, Loss of Offsite Power with Safety Injection - D/G 2B-B Test, Revision 43
* 2-SI-OPS-088-001.0, Phase A Isolation Test, Revision 20
* 2-SI-OPS-082-026.A, Loss of Offsite Power with Safety Injection - D/G 2A-A Test, Revision 47
* 2-SI-OPS-000-009.0, Actuation of ECCS and Boron Injection Flow path valves via Safety Injection Signal, Revision 9  Ice Condenser Surveillance Test:
* 0-SI-MIN-061-107.0, Ice Condenser Floor Drains, Revision 2
* 0-SI-MIN-061-109.0, Ice Condenser Intermediate and Lower Inlet Doors and Vent Curtains, Revision 5 Containment Isolation Valve (CIV) Surveillance Tests:
* 0-SI-SLT-062-258.1, Containment Isolation Valve Local Leak Rate Test Chemical and Volume Control System, Revision 11
* 2-SI-OPS-088-003.0, Phase B Containment Isolation Test, Revision 10
 
====b. Findings====
No findings were identified.
 
===Cornerstone:===
Emergency Preparedness  1EP6 Drill Evaluation
 
====a. Inspection Scope====
Resident inspectors evaluated the conduct of a routine licensee emergency drill on    April 1, 2014, to identify any weaknesses and deficiencies in classification, notification, and protective action recommendation (PAR) development activities. This drill involved beyond design basis events and utilized the licensee's severe accident mitigation guidelines (SAMG). The inspectors evaluated emergency response operations in the simulated control room, as well as the technical support center, to verify that event classification and notifications were performed in accordance with EPIP-1, Emergency Plan Classification Matrix, Revision 51. The inspectors verified that the licensee properly utilized the SAMGs. The inspectors also attended the licensee critique of the drill to compare any inspector observed weakness with those identified by the licensee in order to verify whether the licensee was properly identifying deficiencies. This activity constituted one inspection sample.
 
====b. Findings====
No findings of significance were identified.
 
==RADIATION SAFETY==
(RS)
Cornerstones:  Occupational Radiation Safety and Public Radiation Safety
{{a|2RS1}}
==2RS1 Radiological Hazard Assessment and Exposure Controls==
 
====a. Inspection Scope====
Hazard Assessment and Instructions to Workers:  During facility tours, the inspectors directly observed labeled radioactive material and postings for radiation areas and High Radiation Areas (HRAs) established within the Radiologically Controlled Area (RCA) of the Unit 2 (U2) upper and lower containments, Auxiliary Building and Dry Active Waste (DAW) Storage Facility. The inspectors independently measured radiation dose rates or directly observed conduct of licensee radiation surveys for selected RCA areas, including the Independent Spent Fuel Storage Installation (ISFSI). The inspectors reviewed and verified survey records for several plant areas including surveys for alpha emitters, airborne radioactivity, and gamma surveys with a range of dose rate gradients. The inspectors reviewed several radiation work permit (RWP) details to assess communication of radiological control requirements and current radiological conditions to workers. The inspectors reviewed selected Electronic Dosimeter (ED) dose and dose rate alarms, to verify workers properly responded to the alarms and that the licensee's review of the events was appropriate. The inspectors observed jobs in radiologically risk-significant areas including HRAs and areas with, or with the potential for airborne activity.
 
Contamination and Radioactive Material Control:  The inspectors observed the release of potentially contaminated items from the RCA and from contaminated areas (i.e., U2 containment). The inspectors also reviewed the procedural requirements for, and equipment used to perform, the radiation surveys for release. During plant walk downs, the inspectors evaluated radioactive material storage areas and containers, including satellite RCAs and yard areas, assessing material condition, posting/labeling, and control of materials/areas. In addition, the inspectors reviewed the sealed source inventory and verified labeling, storage conditions, and leak testing of selected sources.
 
Radiological Hazards Control and Work Coverage:  The inspectors evaluated licensee performance in controlling worker access to radiologically significant areas and monitoring jobs in-progress during the week of the onsite inspection. The inspectors also reviewed the procedural guidance for multi and extremity badging. For HRA tasks involving significant dose rate gradients, the inspectors evaluated the use and placement of whole body and extremity dosimetry to monitor worker exposure. The inspectors reviewed RWPs for use in airborne areas, ensuring the prescribed controls were appropriate for the conditions as identified in radiological surveys and air samples. ED alarm set points and worker stay times were evaluated against area radiation survey results for containment and auxiliary building activities.
 
Risk Significant High Radiation Areas and Very High Radiation Area Controls:  The inspectors evaluated access barrier effectiveness for selected Locked High Radiation Area (LHRA) and Very High Radiation Area (VHRA) locations. Changes to procedural guidance for LHRA and VHRA controls were discussed with Radiation Protection (RP)supervisors. During plant walk downs of the U2 Containment and Auxiliary Building, the inspectors verified the posting/locking of LHRA/VHRA areas. Established radiological controls (including airborne controls) were evaluated for selected tasks including work in auxiliary building HRAs, and radiological waste processing and storage. In addition, licensee controls for areas where dose rates could change significantly as a result of plant shutdown and refueling operations were reviewed and discussed.
 
Radiation Worker Performance and RP Technician Proficiency:  The inspectors observed radiation worker performance through direct observation. Jobs observed included routine waste packaging activities in the auxiliary building and routine survey activities in the Auxiliary Building and Upper and Lower Containments in high radiation and contaminated areas. The inspectors also observed health physics technicians (HPTs) providing pre-job/RWP briefings, releasing material from the RCA, and providing field coverage of jobs. Occupational workers' adherence to selected RWPs and HPT proficiency in providing job coverage were evaluated through direct observations and interviews with licensee staff. ED alarm set points and worker stay times were evaluated against area radiation survey results for reviewed RWPs.
 
Problem Identification and Resolution:  PERs associated with radiological hazard assessment and control were reviewed and assessed. The inspectors evaluated the licensee's ability to identify, characterize, prioritize, and resolve the identified issues in accordance with procedure NPG-SPP-22-300, Corrective Action Program, (CAP)
Revision (Rev.) 1. The inspectors also evaluated the scope of the licensee's internal audit program and reviewed recent assessment results. RP activities were evaluated against the requirements of Updated Final Safety Analysis Report (UFSAR) Section 12; Technical Specifications (TS) Sections 6.12; 10 CFR Parts 19 and 20; and approved licensee procedures. Licensee programs for monitoring materials and personnel released from the RCA were evaluated against 10 CFR Part 20 and IE Circular 81-07, Control of Radioactively Contaminated Material. Documents reviewed are listed in the Attachment.
 
====b. Findings====
 
=====Introduction:=====
The inspectors identified a Green, self-revealing, Non-cited Violation (NCV) of TS 6.12.1, "High Radiation Area," for two examples of individuals entering an HRA without meeting the entry requirements as specified in TS 6.12.1.b and e.
 
=====Description:=====
On May 16, 2014, with the U2 reactor shutdown for refueling, a contract employee who was staging equipment and two contract decontamination technicians, working on separate jobs in the upper reactor containment, entered the same posted HRA near the reactor cavity. One of the decontamination technicians and the contractor staging equipment received dose rate alarms shortly after crossing the HRA boundary.
 
Upon receiving the alarms, both individuals exited the area and contacted health physics (HP) as required. The two decontamination technicians were on RWP Number (No.) 210061 with a dose setpoint of 31 mrem and dose rate setpoint of 91 milli-rem per hour (mrem/hr). The decontamination worker's ED indicated a maximum dose rate of 97 mrem/hr. The worker staging equipment was on RWP No. 240051 with a dose setpoint of 21 mrem and dose rate setpoint of 81 mrem/hr. That worker's ED indicated a maximum dose rate of 133 mrem/hr. Accessible general area dose rates based on surveys in the area near the time of the event were as high as 160 mrem/hr at 30 centimeters (cm).
 
In both cases the workers had only been briefed for entry into "Radiation Areas" in the upper reactor containment and that expected dose rates in this area were 3-10 mrem/hr. They were not wearing the prescribed alarming dosimetry for an HRA entry, were not on an RWP that allowed HRA entry, and were not knowledgeable of the actual dose rates in the area. The licensee entered these events into their CAP (PERs 886668 and 886160).
 
Immediate corrective actions included restricting worker access to the RCA and issuance of communications to the site and within the RP organization to reinforce roles in RWP adherence and access control.
 
=====Analysis:=====
The inspectors determined that entry into a HRA without meeting the entry requirements specified in TS 6.12.1 was a performance deficiency. This finding is more than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute of Human Performance and adversely affects the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Workers permitted entry into HRAs with inadequate knowledge of actual radiological conditions could receive unintended occupational exposures. The finding was evaluated using the Occupational Radiation Safety Significance Determination Process (SDP). The finding was not related to ALARA planning, nor did it involve an overexposure or substantial potential for overexposure, and the ability to assess dose was not compromised.
 
Therefore, the inspectors determined the finding to be of very low safety significance (Green). The inspectors noted that the workers responded properly to the ED dose rate alarms thereby limiting their potential for unintended exposure. This finding involved the cross-cutting aspect of Human Performance, Avoid Complacency [H.12] because workers failed to apply appropriate error reduction tools while participating in pre-job briefs and prior to crossing the HRA boundaries.
 
=====Enforcement:=====
TS 6.12.1, "High Radiation Area", requires in part, entries into HRAs be controlled by means of an RWP, associated radiation surveys, and other appropriate radiation protection equipment and measures and except for individuals qualified in RP procedures or escorted by such individuals, entry into such areas shall only be made after dose rates in the area have been determined and entry personnel are made knowledgeable of them. Contrary to the above, on May 16, 2014, workers entered a HRA using an RWP that did not allow HRA access, without using the proper alarming dosimetry, and without knowledge of the actual dose rates in the area. Because this violation was of very low safety significance and it was entered into the licensee's CAP (PERs 886668 and 886160), this violation is being treated as an NCV, consistent with the Enforcement Policy: NCV 05000327/328, 2014003-02, Failure to Comply with Entry requirements to a HRA.
{{a|2RS2}}
==2RS2 Occupational ALARA Planning and Controls==
 
====a. Inspection Scope====
Work Planning and Exposure Tracking:  The inspectors reviewed work activities and their collective exposure estimates associated with the previous Unit 1 (U1) refueling outage, as well as the current U2 refueling outage 19 (U2R19). The U1 refueling outage 19 (U1R19) and U2R19 ALARA planning packages (ALARA Plans) were reviewed for the following high collective exposure tasks:  Refueling operations, Mechanical Maintenance, Plant Services, RP and Modifications. For the selected tasks, the inspectors reviewed the assumptions and basis for the dose rate and man-hour estimates. The inspectors discussed with ALARA staff the means by which wrench-hours were derived from the work order hours provided by craft supervision to ALARA staff. The inspectors verified the licensee had established several means to track and trend doses for ongoing work activities. The inspectors evaluated the incorporation of exposure reduction initiatives and operating experience, including historical post-job reviews, into RWP requirements. Collective dose data for selected tasks were compared with established dose estimates and evaluated against procedural criteria (trigger points) for additional ALARA review. Where applicable, changes to established estimates were discussed with ALARA planners and evaluated against work scope changes or unanticipated elevated dose rate. The inspectors discussed the operation of the Station ALARA Committee with the Site Vice President, the RP Manager and the ALARA Health Physicist. For ALARA Plans from U1R19, the inspectors compared the results achieved in terms of actual dose versus (vs.) planned dose and actual hours vs.
 
estimated hours, reviewed in-progress and post-job ALARA reviews, and discussed the job planning, performance, and reviews with ALARA staff. For ALARA Plans associated with U2R19, the inspectors reviewed dose-to-date on select jobs, comparing estimates with actuals, and observed development of selected in-progress reviews.
 
Source Term Reduction and Control:  The inspectors reviewed the collective exposure three-year rolling average (TYRA) from 2011 - 2013 and reviewed historical outage collective exposure trends. Through interviews with licensee staff and document review, the inspectors assessed the licensee's current activities related to source term reduction, including elevated zinc injection on U2, on-line chemistry using pH 7.4 to minimize corrosion product transport, extended reactor coolant pump run time to allow better cleanup during shutdown, ultrasonic fuel cleaning, and response to fuel defects during previous operating cycles. The inspectors discussed the unexpectedly high activity of shutdown crud burst and changes expected in the short and long term relative abundances of Cobalt-58 and Cobalt-60 that would result from the change in the steam generator tube alloys and increasing the number of steam generator tubes by about a third. The dose implications of the various cobalt reduction activities coupled to the change in tube alloys for the next few outages was also discussed  Radiation Worker Performance:  Radiation worker performance was also observed and evaluated as part of Inspection Procedure 71124.01 and is documented in section 2RS1. While observing job tasks, the inspectors evaluated the use of remote technologies to reduce dose including teledosimetry and remote visual monitoring. Jobs observed were associated with the refueling and maintenance outage.
 
Problem Identification & Resolution:  Licensee CAP documents associated with ALARA planning and controls were reviewed and assessed. This included a review of selected Action Requests (PERs), self-assessments, and audits. The inspectors evaluated the licensee's ability to identify, characterize, prioritize, and resolve the identified issues in accordance with procedure NPG-SPP-22.300, Corrective Action Program, Rev. 1. The inspectors also evaluated the scope and frequency of the licensee's self-assessment program and reviewed recent assessment results.
 
ALARA program activities were evaluated against the requirements of UFSAR Section 12, Radiation Protection; TS Section 6.8, Procedures and Programs; 10 CFR Part 20; and approved licensee procedures. Documents reviewed are listed in the Attachment.
 
====b. Findings====
No findings were identified.
 
{{a|2RS3}}
==2RS3 In-Plant Airborne Radioactivity Control and Mitigation==
 
====a. Inspection Scope====
Engineering Controls:  The inspectors reviewed the use of temporary and permanent engineering controls to mitigate airborne radioactivity during U2R19 for steam generator (S/G) and U2 Thimble eddy current testing and the DAW Storage Building. The use of the U2 Containment Purge to minimize airborne concentrations in containment during refuel activities was discussed with licensee personnel. In addition, inspectors observed the placement and use of high efficiency particulate air negative pressure units, and air sampling equipment during observations of jobs in-progress.
 
Use of Respiratory Protection Devices & Self-Contained Breathing Apparatus for Emergency Use:  Inspectors reviewed the use of respiratory protection devices to limit the intake of radioactive material, including devices used for routine tasks and devices stored for use in emergency situations. Inspectors observed the physical condition of Self-Contained Breathing Apparatus (SCBA) units, negative pressure respirators (NPR)s, powered air purifying respirators and device components staged for routine and emergency use throughout the plant. SCBA bottle air pressure, the number of units, and the number of spare masks and air bottles available was also evaluated by inspectors. The inspectors reviewed maintenance records for selected SCBA units for the past year and evaluated SCBA and NPR compliance with National Institute for Occupational Safety and Health certification requirements. The inspectors also reviewed records of Grade D (or better) air quality testing for supplied-air devices and SCBA bottles. In addition, the inspectors walked-down the compressor used for filling SCBA bottles. The inspectors reviewed the status and surveillance records of SCBAs staged for in-plant use during emergencies through review of records and walk-down of SCBA staged in the control room and selected locations.
 
The inspectors verified the licensee had procedures in place to ensure that the use of respiratory protection equipment was ALARA when engineering controls were not practicable. Control room operators and fire brigade were interviewed on the use of the devices including SCBA bottle change-out and use of corrective lens inserts. Respirator qualification records were reviewed and cross checked for several control room operators. In addition, qualifications for individuals responsible for testing and repairing SCBA vital components were evaluated through review of training records. Selected maintenance records for SCBA units and air cylinder hydrostatic testing documentation were reviewed.
 
The inspectors verified that the licensee has procedural requirements in place for evaluating air samples for the presence of alpha emitters and reviewed airborne radioactivity and contamination survey records for selected plant areas to ensure air samples are screened and evaluated per the procedure requirements.
 
The inspectors walked-down the respirator issue and storage locations and verified that the equipment was appropriately stored and maintained. Records of monthly and quarterly inventory and inspection of the equipment were also reviewed by the inspectors. The inspectors discussed the process for issuing respirators, and verified that selected individuals qualified for respirator and/or SCBA use had completed the required training, fit-test, and medical evaluation.
 
Problem Identification and Resolution:  Licensee CAP documents associated with the control and mitigation of in-plant radioactivity were reviewed and assessed. This included review of selected PERs related to use of respiratory protection devices including SCBA. The inspectors evaluated the licensee's ability to identify, characterize, prioritize, and resolve the identified issues in accordance with procedure NPG-SPP-22-300, Corrective Action Program, Rev.1. The inspectors also evaluated the scope of the licensee's internal audit program and reviewed recent assessment results.
 
RP activities were evaluated against the requirements UFSAR Section 12; 10 CFR Parts 19 and 20; and approved licensee procedures. Documents and records reviewed are listed in the Attachment.
 
====b. Findings====
No findings were identified.
 
{{a|2RS4}}
==2RS4 Occupational Dose Assessment==
 
====a. Inspection Scope====
External Dosimetry:  The inspectors reviewed National Voluntary Laboratory Accreditation Program certification data and discussed program guidance for storage, processing, and evaluation of results for active and passive personnel dosimeters currently in use. Comparisons between ED and thermo-luminescent dosimeter data were discussed in detail. The inspectors reviewed ED alarm logs and reviewed licensee's dosimeter incident reports and assessment actions for selected alarm events.
 
Internal Dosimetry:  Program guidance and assessment results for internally deposited radionuclides were reviewed. The inspectors reviewed selected Whole Body Count (in vivo) analyses from September 2012 to May 2014 as well as in-vitro assessments of tritium exposures to workers entering Unit 2 containment at power during this period.
 
The licensee's methods used in these assessments as well as the programs for collection and analysis of special bioassay samples were discussed with licensee staff. Special Dosimetric Situations:  The inspectors evaluated the licensee's use of multi-badging, extremity dosimetry, and dosimeter relocation within non-uniform dose rate fields and reviewed assessments for U2R19 for S/G maintenance workers. Worker monitoring in neutron areas was discussed with licensee staff. The inspectors also reviewed records of monitoring for declared pregnant workers from September 2012 to May 2014 and discussed monitoring guidance with dosimetry staff. In addition, methods for shallow dose assessments were reviewed and discussed.
 
Problem Identification and Resolution:  The inspectors reviewed and discussed selected CAP documents associated with occupational dose assessment. The inspectors evaluated the licensee's ability to identify and resolve the issues in accordance with procedure NPG-SPP-22-300, Corrective Action Program, Rev.1. The inspectors also discussed the scope of the licensee's internal audit program and reviewed recent assessment results.
 
Occupational dose assessment activities were evaluated against the requirements of UFSAR Section 12; TS Section 6; 10 CFR Parts 19 and 20; and approved licensee procedures. Documents reviewed are listed in the Attachment.
 
====b. Findings====
No findings were identified.
 
{{a|2RS5}}
==2RS5 Radiation Monitoring Instrumentation==
 
====a. Inspection Scope====
Radiation Monitoring Instrumentation:  During walk-downs of the auxiliary building and the RCA exit point, the inspectors observed installed radiation detection equipment.
 
These included area radiation monitors (ARMs), liquid and gaseous effluent monitors, personnel contamination monitors (PCMs), small article monitors (SAMs), and portal monitors (PMs). The inspectors observed the physical location of the components and noted their material condition.
 
In addition to equipment walk-downs, the inspectors reviewed source checks of various portable and fixed detection instruments, including ion chambers, teletectors, PCMs, SAMs, PMs, and an iSOLO alpha/beta counting system. The inspectors reviewed calibration records and evaluated alarm set-point values for PCMs, PMs, effluent monitors, an ARM, and a SAM. This included a sampling of instruments used for post-accident monitoring such as a containment high-range radiation monitor and effluent monitors for noble gas and iodine. The radioactive source used to calibrate an effluent monitor was evaluated for traceability to national standards. Calibration stickers on portable survey instruments were noted during inspection of the storage area for ready-to-use equipment. The most recent 10 CFR Part 61 analysis for DAW was reviewed to determine if calibration and check sources are representative of the plant source term.
 
The inspectors also reviewed count room calibration records for a gamma spectroscopy germanium detector and a liquid scintillation detector.
 
Effectiveness and reliability of selected radiation detection instruments were reviewed against details documented in the following:  10 CFR Part 20; NUREG-0737, Clarification of TMI Action Plan Requirements; UFSAR Chapters 11 and 12; and applicable licensee procedures.
 
Problem Identification and Resolution:  The inspectors reviewed selected PER reports in the area of radiological instrumentation. The inspectors evaluated the licensee's ability to identify and resolve the issues in accordance with procedure NPG-SPP-22.300, Corrective Action Program, Rev. 1. Documents and records reviewed are listed in the
.
 
====b. Findings====
No findings were identified.


==OTHER ACTIVITIES==
                                            5
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Security 
.2    Complete System Walkdown
{{a|4OA1}}
  a. Inspection Scope
==4OA1 Performance Indicator (PI) Verification==
      The inspectors performed a complete system walk down of the Unit 2 Main Steam and
      support systems to verify proper equipment alignment, to identify any discrepancies that
      could impact the function of the system and increase risk, and to verify that the licensee
      properly identified and resolved equipment alignment problems that could cause events
      or impact the functional capability of the system.
      The inspectors reviewed the Updated Final Safety Analysis Report (UFSAR), system
      procedures, system drawings, and system design documents to determine the correct
      lineup and then examined system components and their configuration to identify any
      discrepancies between the existing system equipment lineup and the correct lineup.
      During the walkdown, the inspectors reviewed the following:
      *    Valves were correctly positioned and did not exhibit leakage that would impact the
          functions of any given valve.
      *    Electrical power was available as required.
      *    Major system components were correctly labeled, lubricated, cooled, ventilated, etc.
      *    Hangers and supports were correctly installed and functional.
      *    Essential support systems were operational.
      *    Ancillary equipment or debris did not interfere with system performance.
      *    Valves were locked as required by the locked valve program.
      *    Major system components were correctly labeled.
      *    Visible cabling appeared to be in good material condition.
      In addition, the inspectors reviewed outstanding maintenance work requests and design
      issues on the system to determine whether any condition described in those work
      requests could adversely impact current system operability. Documents reviewed are
      listed in the Attachment. This activity constituted one inspection sample.
  b. Findings
      No findings were identified.
1R05 Fire Protection
.1    Fire Protection Tours
  a. Inspection Scope
      The inspectors conducted a tour of the five areas important to safety listed below to
      assess the material condition and operational status of fire protection features. The
      inspectors evaluated whether: combustibles and ignition sources were controlled in
      accordance with the licensees administrative procedures; fire detection and suppression
      equipment was available for use; passive fire barriers were maintained in good material
                                                                                      Enclosure


====a. Inspection Scope====
                                          6
The inspectors sampled licensee submittals for the five PIs listed below for the period from January 2013 through March 2014 for both Unit 1 and Unit 2. Definitions and guidance contained in Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Indicator Guideline, Revision 6, were used to determine the reporting basis for each data element in order to verify the accuracy of the PI data reported during that period.
    condition; and compensatory measures for out-of-service, degraded, or inoperable fire
    protection equipment were implemented in accordance with the licensees fire plan.
    Documents reviewed are listed in the Attachment. This activity constituted five
    inspection samples.
    *  Control Building Elevation 669 (Mechanical Equipment Room, 250 VDC Battery and
        Battery Board Rooms)
    *  Control Building Elevation 685 (Auxiliary Instrument Rooms)
    *  Turbine Building Elevation 706
    *  Control Building Elevation 706 (Cable Spreading Room)
    *  Control Building Elevation 732 (Mechanical Equipment Room and Relay Room)
b.  Findings
    No findings were identified.
1R06 Flood Protection Measures
    Annual Review of Cables Located in Underground Bunkers/Manholes
a. Inspection Scope
    The inspectors conducted a review of licensee inspections of safety-related cables
    located in underground bunkers/manholes subject to flooding. Specifically, inspectors
    reviewed maintenance records of inspections for the previous 12 months to determine if
    water was present and, if found, whether it would affect safety-related system operation.
    In addition, the inspectors reviewed the licensees corrective action program (CAP) to
    ensure that the licensee was identifying underground cabling issues and that they were
    properly addressed for resolution. Documents reviewed are listed in the Attachment.
    This activity constituted one inspection sample.
b.  Findings
    No findings were identified.
1R08 Inservice Inspection Activities
a.  Inspection Scope
    Non-Destructive Examination Activities and Welding Activities: From May 19, 2014,
    through May 30, 2014, the inspectors conducted an onsite review of the implementation
    of the licensees in-service inspection (ISI) program for monitoring degradation of the
    reactor coolant system (RCS), risk-significant piping and components, and containment
    systems in Unit 2. The inspectors activities included a review of selected samples of
    non-destructive examinations (NDE) to evaluate compliance with the applicable edition
    of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel
                                                                                      Enclosure


===Cornerstone:===
                                      7
Initiating Events
Code (BPVC), Section XI, and to verify that indications and defects (if present) were
* Unplanned Scrams per 7000 Critical Hours
appropriately evaluated and dispositioned, in accordance with the requirements of the
* Unplanned Scrams with Complications
ASME Section XI acceptance standards.
* Unplanned Power Changes per 7000 Critical Hours  The inspectors reviewed selected Licensee Event Reports (LERs) and portions of operator logs to verify whether the licensee had accurately identified the number of scrams and unplanned power changes that occurred during the previous four quarters for both units. The inspectors also reviewed the accuracy of the number of critical hours reported and the licensee's basis for addressing the criteria for complications for each of the reported scrams. Documents reviewed are listed in the Attachment. Cornerstone:  Occupational Radiation
The inspectors directly observed the following NDE, mandated by the ASME Code, to
* Occupational Exposure Control Effectiveness  The inspectors reviewed PI data collected from November 2013 through May 2014, for the Occupational Exposure Control Effectiveness PI. For the reviewed period, the inspectors assessed PER records to determine whether HRA, VHRA or unplanned exposures, resulting in TS or 10 CFR 20 non-conformances, had occurred during the review period. The inspectors reviewed RCA exit transactions with exposures in excess of 100 milli-rem in order to determine compliance with the requirements of the RWP.
evaluate compliance with the ASME Code Section XI, and Section V requirements, and
if any indications or defects were detected, to evaluate if they were dispositioned in
accordance with the ASME Code, or an NRC-approved alternative requirement.
*    Visual Examination (VT) - 3, Steam Generator (SG) Upper Lateral Restraint
    SGH-4-1, ASME Code Class 2
*    General Visual Examination, Containment Moisture Barrier, Examination Category
    E-A, Item No E1.30
Inspectors directly observed the calibration of Ultrasonic Test (UT) equipment, and later
reviewed UT examination results for welds associated with a feedwater elbow
attachment to the SG safe end.
*   Equipment Calibration for UT, Feedwater Piping Weld FDF-011A, ASME Code
    Class 2, Augmented Inspection
*    Equipment Calibration for UT, Feedwater Piping Weld FDF-010C, ASME Code
    Class 2, Augmented Inspection
The inspectors reviewed records of the following NDE inspections and methods
mandated by the ASME Code Section XI or augmented inspections, in order to evaluate
compliance with the ASME Code Section XI and Section V requirements, and if any
indications and defects were detected, to evaluate if they were dispositioned in
accordance with the ASME Code or NRC-approved alternative requirements.
*    VT-3, Pipe Support, 2-CVCH-585, ASME Code Class 2
*    VT-3, Pipe Support, 2-CVCH-584, ASME Code Class 2
*    VT-3, Pipe Support, 2-CVCH-586, ASME Code Class 2
*    Penetrant Testing (PT), Integral Attachment Weld, 2-CVCH-006-IA, Code Class 1
*    UT, Feedwater Piping Weld FDF-011A, ASME Code Class 2, (Augmented)
*    UT, Feedwater Piping Weld FDF-010C, ASME Code Class 2, (Augmented)
The inspectors reviewed the following surface examination records with recordable
indications that were analytically evaluated and accepted for continued service, against
the ASME Code Section XI, or an NRC-approved alternative.
*    PT, Lug to Pipe Weld, 2-SIH-020-IA, ASME Code Class 1
                                                                                  Enclosure


The reviewed data were assessed against guidance contained in Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment Indicator Guideline," Rev. 6.
                                      8
No ASME Class 1, 2, or 3 welding activities were in progress during the NRC ISI
inspector site visit. Therefore, the inspectors reviewed the previously completed welding
activity work order (WO), referenced below, in order to evaluate compliance with the
intent of procedures, and the ASME Code. Specifically, the inspectors reviewed the WO
package, the WO VT-2 leakage examination requirements and results.
*    WO No. 112354373, SQN-2-VLV-0012-0817, Valve Replacement, ASME Class 2
Pressurized Water Reactor Vessel Upper Head Penetration Inspection Activities: For
the Unit 2 reactor vessel head, a full bare metal visual (BMV) examination was not
required this outage pursuant to 10 CFR 50.55a. Therefore, no reviews were conducted
for this inspection attribute. A volumetric examination of the Unit 2 vessel upper head
penetration (VUHP) was required this outage. Therefore, inspectors observed and
reviewed a sample of the Unit 2 UT examination results, which included NDE reports for
VUHP Nos. 53, 56, and 60. The inspectors also performed a comparison of the current
UT results to the previous UT examination results for the sample penetrations. These
comparisons were used to determine if the activities, including the disposition of
indications and defects, were conducted in accordance with the requirements of ASME
Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D). In particular, the inspectors
evaluated if the required UT examination scope/coverage was achieved and limitations
(if applicable) were recorded in accordance with the licensee procedures. The licensee
did not identify any relevant indications that needed to be accepted by analysis for
continued service during the UT examination. Additionally, the licensee did not perform
any welding repairs to the vessel head penetrations since the beginning of the last Unit 2
refueling outage; therefore, no NRC review was completed for these inspection
procedure attributes.
Boric Acid Corrosion Control Inspection Activities: The inspectors reviewed the
licensees boric acid corrosion control (BACC) program activities, to ensure
implementation with commitments made in response to NRC Generic Letter 88-05,
Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary, and applicable
industry guidance documents. Specifically, the inspectors performed an onsite record
review of procedures and the results of the licensees containment walkdown inspections
performed during the current spring refueling outage. The inspectors also interviewed
the BACC program owner, conducted an independent walkdown of two levels of
containment, in order to evaluate compliance with the licensees BACC program
requirements, and verify that degraded or non-conforming conditions, such as boric acid
leaks, were properly identified and corrected in accordance with the licensees BACC
and corrective action program (CAP).
The inspectors reviewed the following problem evaluation report (PER), and associated
corrective actions related to evidence of boric acid leakage, to evaluate if the corrective
actions completed were consistent with the requirements of the ASME Code Section XI,
and 10 CFR Part 50, Appendix B, Criterion XVI, and Industry Guidelines.
*    PER 743110, Degraded Non-conforming condition for PDO on RCS leakage and Ice
    on Intermediate Deck Doors (IDD), 7/19/13
                                                                                  Enclosure


===Cornerstone:===
                                      9
Public Radiation Safety
The inspectors reviewed the following engineering evaluations completed for evidence of
* Radiological Control Effluent Release Occurrences  The inspectors reviewed the Radiological Control Effluent Release Occurrences PI results for the Public Radiation Safety Cornerstone from November 2013, through May 2014. For the assessment period, the inspectors reviewed cumulative and projected doses to the public and PER documents related to Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual issues.
boric acid leakage to determine if degraded components were documented in the CAP.
The inspectors also reviewed corrective actions for any degraded components to
determine if they met the applicable requirements of the ASME Code, Section XI, and/or
NRC-approved alternatives.
*  PER 888330, Boric Acid Leakage Evaluation, Reactor Cavity Nozzle Cover Seal
    Leaking into Keyway, 6/24/14
*  PER 890230, Evaluation of Boric Acid Corrosion Damage, 2-SNUB-068-RCH30,
    6/7/14
*   SR 889942, Determine Available Margins in Pipe Support Attributes, 2-RCH-0028,
    5/26/14
Steam Generator Tube Inspection Activities: The inspectors reviewed the eddy current
(EC) examination activities performed in Unit 2 SGs 1, 2, 3, and 4 during the end-of-
cycle 19 refueling outage, to verify compliance with the licensees Technical
Specifications, ASME BPVC Section XI, and Nuclear Energy Institute (NEI) 97-06,
Steam Generator Program Guidelines. The inspectors interviewed licensee personnel
and vendor staff responsible for the SG inspection project, and reviewed documentation
associated with the SG inspections and integrity assessments, as described in this
report section.
The inspectors reviewed the scope of the EC examinations to verify that known and
potential areas of tube degradation were inspected. The inspectors also verified that
inspection scope expansion criteria were implemented based on inspection results, as
directed by the Electric Power Research Institute (EPRI) Pressurized Water Reactor
Steam Generator Examination Guidelines, Revision 7.
The inspectors reviewed documentation for a sample of EC data analysts, EC probes,
and EC testers to verify that personnel and equipment were qualified to detect the
existing and potential degradation mechanisms applicable to Sequoyahs SG tubes, in
accordance with the EPRI Examination Guidelines. This review included a sample of
site-specific Examination Technique Specification Sheets (ETSSs) that were selected
based on plant-specific and industry operating experience, to ensure that their
qualification and site-specific implementation were consistent with Appendix H or I of the
EPRI Examination Guidelines. The selected ETSSs for review consisted of bobbin and
rotating probe techniques that were used to detect wear at the tube interface with
support structures (i.e., tube support plates, anti-vibration bar (AVB), and flow
distribution baffle plate), and wear associated with foreign objects.
The inspectors also reviewed a sample of EC data with a qualified data analyst to
confirm that data analysis was performed in accordance with the applicable ETSSs and
site-specific analysis guidelines. The inspectors verified that the equipment
configuration was consistent with the essential parameters of the applicable technique.
                                                                                  Enclosure


====b. Findings====
                                    10
No findings were identified.
The inspectors also verified that recordable indications were detected and sized in
accordance with vendor procedures. As part of the EC data review, the inspectors
verified that the EC indications on each selected tube were consistent with historical
data relative to the number of indications, location, and size. The sample of EC data
selected for review is listed below:
      Steam            Tube          Eddy Current            Indication Type
    Generator      Row/Column            Probe
        2            R93/C59            Bobbin                  AVB wear
        2            R93/C59        MRPC + point              AVB Wear
        2            R93/C59            Array                  AVB Wear
        2            R89/C59            Bobbin                  AVB Wear
        2            R89/C59        MRPC + point                AVB Wear
        4            R93/C47            Bobbin              Proximity Signal
        4            R93/C47            Array              Proximity Signal
        2            R5/C101            Bobbin          Distorted Support Signal
        2            R5/C101        MRPC + point        Distorted Support Signal
The inspectors selected a sample of wear degradation mechanisms from the Steam
Generator Degradation Assessment, and verified that the in-situ pressure testing criteria
were determined, in accordance with the EPRI Tube Integrity Guidelines. Additionally,
the inspectors reviewed EC indication reports to determine whether tubes with relevant
indications were appropriately screened for in-situ pressure testing.
The inspectors compared the recent EC examination results with the last Operational
Assessment report for SGs to assess the licensees prediction capability for maximum
tube degradation, and number of tubes with indications. The inspectors verified that the
licensees evaluation was conservative and that current examination results were bound
by the Operational Assessment projections.
The inspectors also compared past examination results discussed in the latest
Degradation Assessment with the recent EC examination results to verify that new
degradation mechanisms, if any, were identified and evaluated before plant startup. The
review of EC examination results included the disposition of potential loose part
indications on the SG secondary side, to verify that corrective actions for evaluating and
retrieving loose parts were consistent with the EPRI Guidelines. The inspectors also
reviewed a sample of primary-to-secondary leakage data for Unit 2 to confirm that
operational leakage in all SGs remained below the action level threshold during the
previous operating cycle.
Based on the review of the final EC examination results for all SGs and interviews with
the licensee, the inspectors confirmed that no EC scope expansion was required, and
none of the SG tubes examined met the criteria for plugging or in-situ pressure testing.
                                                                                Enclosure


{{a|4OA2}}
                                          11
==4OA2 Problem Identification and Resolution==
  Furthermore, the inspectors interviewed licensee staff and reviewed a sample of
  secondary side visual inspection results for the SGs 1, 2, 3, and 4 upper bowl areas, to
  verify that potential areas of degradation based on site-specific operating experience
  were inspected, and appropriate corrective actions were taken to address degradation
  indications. This review included the results of Foreign Object Search and Retrieval
  (FOSAR) activities in all SGs, and an evaluation for loose parts in the secondary side of
  SGs 1, 2, 3, and 4.
  Identification and Resolution of Problems: The inspectors reviewed a sample of ISI-
  related problems which were identified by the licensee, and entered into the CAP as
  PERs. The inspectors reviewed the PERs to confirm that the licensee had appropriately
  described the scope of the problem, and had initiated corrective actions. The review
  also included the licensees consideration and assessment of operating experience
  events applicable to the plant. The inspectors performed this review to ensure
  compliance with 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action
  requirements. Documents reviewed are listed in the Attachment.
b. Findings
  Introduction: An NRC-identified Green NCV of 10 CFR 50.55a (g)(4), Inservice
  Inspection Requirements was identified for the licensees failure to perform visual
  examinations of the control rod drive mechanism (CRDM), ASME Class 1, seismic plate
  supports as required by the ASME Code, Section XI.
  Description: The Sequoyah Unit 2 ISI program for the current interval (third interval) was
  prepared in accordance with the 2001 Edition of the ASME Section XI Code, with
  addenda through 2003. Article IWF-2000 of ASME Section XI, Table IWF-2500-1,
  Examination Category Item Number F1.40, requires, in part, periodic VT-3 visual
  examinations of supports (other than piping supports) in Class 1 components. The
  examinations provide reasonable assurance that the supports can continue to perform
  their intended function.
  The CRDM assemblies are ASME Class 1 pressure retaining components that contain a
  series of seismic plate supports to ensure that the allowable design stress limits for the
  CRDM assemblies are not exceeded during a seismic event, which in turn provides
  reasonable assurance that the RCS pressure boundary and control rod function is
  maintained.
  The inspectors identified that the Sequoyah Unit 2 ISI program did not meet the
  requirements of ASME Section XI in that the Class 1 CRDM seismic plate supports, and
  associated load path components, which meet the examination category F1.40, were not
  included in the scope of the program for the first, second, and third ISI intervals. The
  inspectors also identified that this issue applied to the Unit 1 ISI Program.
  The licensee entered this issue into their CAP as PER 889400. The licensee developed
  an operability evaluation and concluded that the supports were operable but non-
  conforming. The evaluation considered previous dimensional verifications of the reactor
  vessel head lift rig components in the area of the CRDM seismic support plates, and as-
                                                                                      Enclosure


===.1 Daily Review===
                                      12
found settings of the seismic plates from a modification project WO package associated
with Unit 1 and 2 cables in the seismic plate area of the lift rig. The WO package
included requirements to insert a gap gauge at each seismic plate screw pad gap to
verify the correct gap was present on Unit 1. The results of Unit 1 as-found gap settings
provided reasonable assurance that the as-found gap settings were adequate for Unit 2
based on the similarities in design, operating conditions, and implementation of outage
maintenance activities. The evaluation also considered that no degradation of the lift rig
intervening steel components in the support load path between the seismic plates and lift
rig struts had been reported in previous outages through the CAP. The licensee also
initiated corrective actions to perform the required visual examinations of the CRDM
seismic plate supports before the end of the current ISI interval in April 2016.
Analysis: Failure to perform the required visual examinations of the CRDM seismic
plates and associated load path components, as required by the ASME Section XI Code,
was a performance deficiency (PD). In accordance with Inspection Manual Chapter
(IMC) 0612 Appendix B, Issue Screening, the PD was more than minor because it was
associated with the protection against external factors attribute of the mitigating systems
cornerstone, and affected the cornerstone objective to ensure availability, reliability, and
capability of systems that respond to initiating events to prevent undesirable
consequence. Specifically, the licensee failed to perform examinations required to
provide reasonable assurance that the support components can perform their intended
function during design basis seismic events, and therefore maintain the stresses in the
CRDM assembly within the allowable design limits, which in turn provides reasonable
assurance that the RCS pressure boundary and control rod function is maintained. The
inspectors screened this finding as Green in accordance with IMC 0609, Appendix A,
Exhibit 2, Mitigating Systems, because the finding did not involve the loss or degradation
of equipment or function specifically designed to mitigate a seismic initiating event. A
crosscutting aspect was not assigned to this finding in accordance with IMC 0612,
Appendix B, because the exclusion of the CRDM seismic plate supports from the scope
of the ISI Program occurred outside of the nominal 3-year period for present
performance, and therefore it was not reflective of present licensee performance.
Enforcement: Title 10 CFR 50.55a(g)(4), Inservice Inspection Requirements, requires
in part that throughout the service life of a boiling or pressurized water-cooled nuclear
power facility, components (including supports) that are classified as ASME Code Class
1, must meet the requirements, except design and access provisions, and preservice
examination requirements set forth in Section XI of editions and addenda of the ASME
BPVC that become effective subsequent to editions specified in paragraphs (g)(2) and
(g)(3) of this Section, and that are incorporated by reference in paragraph (b) of this
Section, to the extent practical within the limitations of design, geometry, and materials
of construction of the components. Section XI of the ASME BPVC, 2001 Edition with
2003 Addenda, Table IWF-2500-1, Examination Category F-A Supports, requires a VT-3
examination of 100 percent of the ASME Class 1 supports, other than piping supports,
every ISI Interval (examination item F1.40), as modified by Notes 1, 2, 3 and 5 of Table
IWF-2500-1.
                                                                                  Enclosure


====a. Inspection Scope====
                                              13
As required by Inspection Procedure 71152, Identification and Resolution of Problems, and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensee's CAP. This was accomplished by reviewing the description of each new PER and attending daily management review committee meetings.
      Contrary to the above, from initial commercial operation until present, the licensee failed
      to perform the required VT-3 examination of ASME Class 1 supports, other than piping
      supports, (i.e., seismic support plates and associated load path components) on the
      CRDM assemblies of Units 1 and 2. The licensee entered the issue into the CAP as
      PER 889400. The licensee initiated corrective actions to perform the required VT-3
      examinations during the next refueling outage in order to restore compliance with the
      10 CFR 50.55a regulations. Because this violation was determined to be of very low
      safety significance (i.e., Green), and the licensee entered the issue in the CAP, this
      violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC
      Enforcement Policy, dated July 9, 2013. This finding will be tracked as NCV 05000327,
      328/2014003-01, Failure to Perform Visual Examination of the Unit 1 and Unit 2 CRDM
      Seismic Plate Supports.
1R11 Licensed Operator Requalification Program
      Quarterly Review
  a. Inspection Scope
      On June 24, 2014, the inspector observed an evaluated simulator scenario and the
      evaluators critique of crew performance. The exercise was performed to provide
      practice to the operating crews in longer duration CPE style scenarios. The inspector
      observed crew performance in terms of: communications; ability to take timely and
      proper actions; prioritizing, interpreting and verifying alarms; correct use and
      implementation of procedures, including the alarm response procedures; timely control
      board operation and manipulation, including high risk operator actions; oversight and
      direction provided by shift manager, including the ability to identify and implement
      appropriate Technical Specification (TS) action; and, group dynamics involved in crew
      performance. The inspector observed the ability of the licensee to administer the
      evaluation and quality of the evaluators critique. The inspector observed scenario
      operations for simulator fidelity to verify that it matched actual plant response. Based on
      crew performance and scenario administration issues, the inspector also reviewed the
      follow-up actions taken to address operator deficiencies and identified administration
      issues. Documents reviewed are listed in the Attachment. This activity constituted one
      inspection sample.
  b. Findings
      No findings were identified
.2    Quarterly Review of Licensed Operator Performance
  a. Inspection Scope
      The inspectors observed and assessed licensed operator performance in the main
      control room during periods of heightened activity or risk. The inspectors reviewed
      various licensee policies and procedures such as OPDP-1, Conduct of Operations,
      NPG-SPP-10.0, Plant Operations, and 0-GO-5, Normal Power Operation. The
                                                                                        Enclosure


====b. Findings and Observations====
                                            14
No findings were identified.
    inspectors utilized activities such as post-maintenance testing, surveillance testing,
    unplanned transients, infrequent plant evolutions, plant startups and shutdowns, reactor
    power and turbine load changes, and refueling and other outage activities to focus on
    the following conduct of operations as appropriate:
    *  operator compliance and use of procedures
    *  control board manipulations
    *  communication between crew members
    *  use and interpretation of plant instruments, indications and alarms
    *  use of human error prevention techniques
    *  documentation of activities, including initials and sign-offs in procedures
    *  supervision of activities, including risk and reactivity management
    *  pre-job briefs
    Specifically, the inspectors observed licensed operator performance during the following
    activities:
    *  Unit 2 reactor shut down and cool down
    *  Unit 2 reactor start up
    Documents reviewed are listed in the Attachment. This activity constituted one
    inspection sample.
b. Findings
    No findings were identified
1R12 Maintenance Effectiveness
a. Inspection Scope
    The inspectors reviewed the maintenance activities, issues, and/or systems listed below
    to verify the effectiveness of the licensees activities in terms of: appropriate work
    practices; identifying and addressing common cause failures; scoping in accordance
    with 10 CFR 50.65(b); characterizing reliability issues for performance; trending key
    parameters for condition monitoring; charging unavailability for performance;
    classification in accordance with 10 CFR 50.65(a)(1) or (a)(2); appropriateness of
    performance criteria for structure, system, or components (SSCs) and functions
    classified as (a)(2); and appropriateness of goals and corrective actions for SSCs and
    functions classified as (a)(1). Documents reviewed are listed in the Attachment. This
    activity constituted one inspection sample.
    *  Cause Determination Evaluation 2741 associated with failure of flow switch (FS) 2-
        FS-74
                                                                                        Enclosure


===.2 Selected Issue Follow-up: Pressurizer Thermal Limit Exceeded===
                                          15
b. Findings
    No findings were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control
a.  Inspection Scope
    The inspectors reviewed the following activities to determine whether appropriate risk
    assessments were performed prior to removing equipment from service for
    maintenance. The inspectors evaluated whether risk assessments were performed as
    required by 10 CFR 50.65(a)(4), and were accurate and complete. When emergent
    work was performed, the inspectors reviewed whether plant risk was promptly
    reassessed and managed. The inspectors also assessed whether the licensees risk
    assessment tool use and risk categories were in accordance with Standard Programs
    and Processes Procedure NPG-SPP-07.1, On-Line Work Management, Revision 3,
    and Instruction 0-TI-DSM-000-007.1, Risk Assessment Guidelines, Revision 9.
    Documents reviewed are listed in the Attachment. The inspectors completed five
    samples.
    *  Unit 1 Yellow probabilistic safety assessment (PSA) risk associated with 1B Residual
        Heat Removal (RHR) pump planned maintenance
    *  emergent work due to failure of Individual Rod Position Indication (IRPI) E-5
    *  maintenance risk review U2R19 Outage Schedule
    *  emergent work due to failure of CS Component Cooling System (CCS) swing pump
        hand switch
    *  emergent work due to failure of Unit 2 vacuum breaker (2-30-573)
b.  Findings
    No findings were identified.
1R15 Operability Evaluations
a.  Inspection Scope
    For the four operability evaluations described in the PERs listed below, the inspectors
    evaluated the technical adequacy of the evaluations to ensure that TS operability was
    properly justified and the subject component or system remained available, such that no
    unrecognized increase in risk occurred. The inspectors compared the operability
    evaluations to UFSAR descriptions to determine if the system or components intended
    function(s) were adversely impacted. In addition, the inspectors reviewed compensatory
    measures implemented to determine whether the compensatory measures worked as
    stated and the measures were adequately controlled. The inspectors also reviewed a
    sampling of PERs to assess whether the licensee was identifying and correcting any
    deficiencies associated with operability evaluations. Documents reviewed are listed in
    the Attachment. This activity constituted four inspection samples.
                                                                                      Enclosure


====a. Inspection Scope====
                                          16
The inspectors performed an in-depth review of PER 809100. On November 9, 2013, during a plant heat-up on Unit 1, the pressurizer thermal limits were exceeded on two separate occurrences, contrary to the Technical Requirements Manual (TRM) Section 3.9.2. Westinghouse performed an analysis of the event and determined that the existing pressurizer design basis analysis bounded this event and there was no increase in the limiting primary stress or the primary-plus-secondary stress range from this event. As a result of this event, the licensee developed an Apparent Cause Analysis (ACE).
    *  PER 855557/864224: Operation Decision Making Instruction (ODMI) for Unit 2
        Power Range Overpower Rod Stop Deviation Alarms
    *  PER 886167: ODMI for Unit 1 Cavity Seal Leakage
    *  PER 855850: Past operability evaluation (POE) associated with 2B RHR 2-FS-74-
        24A failure
    *  PER 897994: Prompt Determination of Operability (PDO) for Unit 2 Turbine Driven
        AFW pump
b.  Findings
    No findings were identified.
1R19 Post-Maintenance Testing
a. Inspection Scope
    The inspectors reviewed the post-maintenance tests associated with the six work orders
    (WOs) listed below to assess whether procedures and test activities ensured system
    operability and functional capability. The inspectors reviewed the licensees test
    procedure to evaluate whether: the procedure adequately tested the safety function(s)
    that may have been affected by the maintenance activity, the acceptance criteria in the
    procedure were consistent with information in the applicable licensing basis and/or
    design basis documents, and the procedure had been properly reviewed and approved.
    The inspectors also witnessed the test or reviewed the test data to determine whether
    test results adequately demonstrated restoration of the affected safety function(s).
    Documents reviewed are listed in the Attachment. This activity constituted six inspection
    samples.
    *  WO 115149300, Rx Vessel Wide Range Level Failed High
    *  WO 115806034, Unit 1 Electric Pulse Repair of IRPI Connectors
    *  WO 114973816, Unit 1 RHR Mini Flow Valve environmental qualification
        maintenance and Inspection
    *  WO 113877775, RHR Return Valve Leak Rate Test for FCV-74-1 and FCV-74-2
    *  WO 113880726, SIS/RHR Hot Leg Check Valve Backseat Test
    *  WO 113875488, Post Maintenance Local Leak Rate Test (as-left) for 2-FCV-63-71,
        2-FCV-63-84, & 2-FCV-63-23
b. Findings
    No findings were identified.
                                                                                    Enclosure


The inspectors reviewed the actions taken to determine if the licensee had adequately addressed the following attributes.
                                              17
* Complete, accurate and timely identification of the problem
1R20 Refueling and Outage Activities
* Evaluation and disposition of operability and reportability issues
    Unit 2 Refueling Outage Cycle 19
* Consideration of previous failures, extent of condition, generic or common cause implications
a.  Inspection Scope
* Prioritization and resolution of the issue commensurate with safety significance
    For the Unit 2 refueling outage that began on May 12, the inspectors evaluated licensee
* Identification of the root cause and contributing causes of the problem
    activities in order to verify that the licensee considered risk in developing outage
* Identification and implementation of corrective actions commensurate with the safety significance of the issue
    schedules, followed risk reduction methods developed to control plant configuration,
    developed mitigation strategies for the loss of key safety functions, and adhered to
    operating license and TS requirements that ensure defense-in-depth. The inspectors
    also walked down portions of Unit 2 not normally accessible during at-power operations
    to verify that safety-related and risk-significant SSCs were maintained in an operable
    condition. Specifically, between May 12 and June 30, the inspectors performed
    inspections and reviews of the following outage activities. This activity constituted one
    inspection sample for Refueling Activities.
    *   Outage Plan. The inspectors reviewed the outage safety plan and contingency plans
        to confirm that the licensee had appropriately considered risk, industry experience,
        and previous site-specific problems in developing and implementing a plan that
        assured maintenance of defense-in-depth.
    *   Reactor Shutdown. The inspectors observed the shutdown in the control room from
        the time the reactor was tripped until operators placed it on the RHR system for
        decay heat removal to verify that TS cool down restrictions were followed. The
        inspectors also toured the lower containment as soon as practicable after reactor
        shutdown to observe the general condition of the reactor coolant system (RCS),
        emergency core cooling system components, and to look for indications of previously
        unidentified leakage inside the polar crane wall.
    *  Licensee Control of Outage Activities. On a daily basis, the inspectors attended the
        licensee outage turnover meeting, reviewed PERs, and reviewed the defense-in-
        depth status sheets to verify that status control was commensurate with the outage
        safety plan and in compliance with the applicable TS when taking equipment out of
        service. The inspectors further toured the main control room and areas of the plant
        daily to ensure that the following key safety functions were maintained in accordance
        with the outage safety plan and TS: electrical power, decay heat removal, spent fuel
        cooling, inventory control, reactivity control, and containment closure. The
        inspectors also observed a tag-out (2-TO-2014-0039, Tag-out of 2B-B Centrifugal
        Charging Pump) to verify that the equipment was appropriately configured to safely
        support the work and testing. To ensure that RCS level instrumentation was properly
        installed and configured to give accurate information, the inspectors reviewed the
        installation of the Mansell level monitoring system. Specifically, the inspectors
        discussed the system with engineering, walked it down to verify that it was installed
        in accordance with procedures and adequately protected from inadvertent damage,
        verified that Mansell indication properly overlapped with pressurizer level instruments
        during pressurizer drain-down, verified that operators properly set level alarms to
        procedurally required set-points, and verified that the system consistently tracked
                                                                                        Enclosure


====b. Findings====
                                            18
There were no findings identified during this review. The inspectors determined that the ACE was probing and involved an extent of condition review, a safety culture analysis, and operating experience review. The ACE also brought to light that the crew that performed the pressurizer heat up did not realize the limit violation. This fact was actually noted by the night shift crew when reviewing data. The ACE also analyzed a near miss that occurred on November 13, 2013. In this instance, the plant was being returned to cold iron conditions due to a pressurizer power operated relief failure and the operators were challenged in maintaining cool-down limits of the pressurizer. Although no limits were exceeded the November 13 incident, the ACE noted several weaknesses in the evolution. The ACE ultimately led to the development of several corrective actions; including procedural changes designed strengthen the operators' awareness of pressurizer pressure control, development of a vendor-performed stress analysis of the event, and addition of the event to the operations training program in order to share information with other crews.
        RCS level while lowering to reduced inventory conditions. The inspectors also
        observed operators compare the Mansell indications with locally-installed ultrasonic
        level indicators during entry into reduced inventory conditions.
    *  Refueling Activities. The inspectors observed fuel movement at the spent fuel pool
        and at the refueling cavity in order to verify compliance with TS and that each
        assembly was properly tracked from core offload to core reload. In order to verify
        proper licensee control of foreign material, the inspectors verified that personnel
        were properly checked before entering any foreign material exclusion (FME) areas,
        reviewed FME procedures, and verified that the licensee followed the procedures.
        To ensure that fuel assemblies were loaded in the core locations specified by the
        design, the inspectors independently reviewed the recording of the licensees final
        core verification.
    *  Reduced Inventory and Mid-Loop Conditions. Prior to the outage, the inspectors
        reviewed the licensees commitments to Generic Letter 88-17. Before entering
        reduced inventory conditions the inspectors verified that these commitments were in
        place, that plant configuration was in accordance with those commitments, and that
        distractions from unexpected conditions or emergent work did not affect operator
        ability to maintain the required reactor vessel level. The inspectors verified that
        licensee procedures for closing the containment upon a loss of decay heat removal
        were in effect, that operators were aware of how to implement the procedures, and
        that other personnel were available to close containment penetrations, if needed. In
        order to reduce outage risk, the licensee elected to not put the plant into mid-loop
        conditions during this particular refueling outage.
    *  Heatup and Startup Activities. The inspectors toured the containment prior to reactor
        startup to verify that debris that could affect the performance of the containment
        sump had not been left in the containment. The inspectors reviewed the licensees
        mode-change checklists to verify that appropriate prerequisites were met prior to
        changing TS modes. Prior to plant startup, the inspectors performed a detailed tour
        of containment to ensure no debris existed that could affect containment sump
        performance given a design basis accident. The inspectors also inspected the
        primary system in containment during Mode 3 with the plant at normal operating
        pressure and temperature in order to verify the leak tightness of the RCS.
b. Findings
    No findings were identified.
1R22 Surveillance Testing
a.  Inspection Scope
    For the 11 surveillance tests identified below, the inspectors assessed whether the
    SSCs involved in these tests satisfied the requirements described in the TS surveillance
    requirements, the UFSAR, applicable licensee procedures, and whether the tests
    demonstrated that the SSCs were capable of performing their intended safety functions.
                                                                                        Enclosure


===.3 Semi-Annual Trend Review===
                                          19
    This was accomplished by witnessing testing and/or reviewing the test data. Documents
    reviewed are listed in the Attachment. This activity constituted 11 inspection samples.
    In-Service Tests:
    *    2-SI-SXP-063-202.0, Safety Injection Pumps 2A-A and 2B-B Comprehensive
          Performance and Check Valve Test, Revision 5
    Routine Surveillance Tests:
    *    0-SI-OPS-092-078.0 Power Range Neutron Flux Channel Calibration by Heat
          Balance Comparison, Revision 23
    *    0-SI-NUC-000-038.0 Unit 2 Shutdown Margin, Revision 75
    *    2-SI-OPS-082-026.B, Loss of Offsite Power with Safety Injection - D/G 2B-B Test,
          Revision 43
    *    2-SI-OPS-088-001.0, Phase A Isolation Test, Revision 20
    *    2-SI-OPS-082-026.A, Loss of Offsite Power with Safety Injection - D/G 2A-A Test,
          Revision 47
    *    2-SI-OPS-000-009.0, Actuation of ECCS and Boron Injection Flow path valves via
          Safety Injection Signal, Revision 9
    Ice Condenser Surveillance Test:
    *    0-SI-MIN-061-107.0, Ice Condenser Floor Drains, Revision 2
    *    0-SI-MIN-061-109.0, Ice Condenser Intermediate and Lower Inlet Doors and Vent
          Curtains, Revision 5
    Containment Isolation Valve (CIV) Surveillance Tests:
    *    0-SI-SLT-062-258.1, Containment Isolation Valve Local Leak Rate Test Chemical
          and Volume Control System, Revision 11
    *    2-SI-OPS-088-003.0, Phase B Containment Isolation Test, Revision 10
b.  Findings
    No findings were identified.
    Cornerstone: Emergency Preparedness
1EP6 Drill Evaluation
  a. Inspection Scope
    Resident inspectors evaluated the conduct of a routine licensee emergency drill on
    April 1, 2014, to identify any weaknesses and deficiencies in classification, notification,
    and protective action recommendation (PAR) development activities. This drill involved
    beyond design basis events and utilized the licensees severe accident mitigation
                                                                                      Enclosure


====a. Inspection Scope====
                                            20
As required by Inspection Procedure 71152, the inspectors performed a semi-annual review of the licensee's corrective action program and associated documents to identify trends that could indicate the existence of a more significant safety issue. The inspectors review was focused on repetitive equipment issues, but also included licensee trending efforts and licensee human performance results. The inspectors review nominally considered the twelve-month period of July 2013 through June 2014, although some examples expanded beyond those dates when the scope of the trend warranted. Specifically, the inspectors considered the results of daily inspector screening discussed in Section 4OA2.1 and reviewed licensee trend reports for the period in order to determine the existence of any adverse trends that the licensee may not have previously identified. This activity constituted one inspection sample for Semi-annual Trend Review.
      guidelines (SAMG). The inspectors evaluated emergency response operations in the
      simulated control room, as well as the technical support center, to verify that event
      classification and notifications were performed in accordance with EPIP-1, Emergency
      Plan Classification Matrix, Revision 51. The inspectors verified that the licensee properly
      utilized the SAMGs. The inspectors also attended the licensee critique of the drill to
      compare any inspector observed weakness with those identified by the licensee in order
      to verify whether the licensee was properly identifying deficiencies. This activity
      constituted one inspection sample.
  b. Findings
      No findings of significance were identified.
2.    RADIATION SAFETY (RS)
      Cornerstones: Occupational Radiation Safety and Public Radiation Safety
2RS1 Radiological Hazard Assessment and Exposure Controls
  a. Inspection Scope
      Hazard Assessment and Instructions to Workers: During facility tours, the inspectors
      directly observed labeled radioactive material and postings for radiation areas and High
      Radiation Areas (HRAs) established within the Radiologically Controlled Area (RCA) of
      the Unit 2 (U2) upper and lower containments, Auxiliary Building and Dry Active Waste
      (DAW) Storage Facility. The inspectors independently measured radiation dose rates or
      directly observed conduct of licensee radiation surveys for selected RCA areas,
      including the Independent Spent Fuel Storage Installation (ISFSI). The inspectors
      reviewed and verified survey records for several plant areas including surveys for alpha
      emitters, airborne radioactivity, and gamma surveys with a range of dose rate gradients.
      The inspectors reviewed several radiation work permit (RWP) details to assess
      communication of radiological control requirements and current radiological conditions to
      workers. The inspectors reviewed selected Electronic Dosimeter (ED) dose and dose
      rate alarms, to verify workers properly responded to the alarms and that the licensees
      review of the events was appropriate. The inspectors observed jobs in radiologically
      risk-significant areas including HRAs and areas with, or with the potential for airborne
      activity.
      Contamination and Radioactive Material Control: The inspectors observed the release
      of potentially contaminated items from the RCA and from contaminated areas (i.e., U2
      containment). The inspectors also reviewed the procedural requirements for, and
      equipment used to perform, the radiation surveys for release. During plant walk downs,
      the inspectors evaluated radioactive material storage areas and containers, including
      satellite RCAs and yard areas, assessing material condition, posting/labeling, and
      control of materials/areas. In addition, the inspectors reviewed the sealed source
      inventory and verified labeling, storage conditions, and leak testing of selected sources.
                                                                                        Enclosure


====b. Findings and Observations====
                                      21
No findings were identified. The inspectors noted a negative trend regarding human performance errors. During the daily reviews, the inspectors noted an increase in human error events. The inspectors then performed a more detailed review of the trend under the semi-annual trend review required by IP 71152. The inspectors concluded there were at least eleven of these events that occurred in the last three months. The inspectors noted this was more than the typical amount of error-related incidents observed during a quarter. The below abbreviated list of PERs involved several human performance related and mis-positioning events as well as procedural non-compliance.
Radiological Hazards Control and Work Coverage: The inspectors evaluated licensee
* PER 868301, EDG 1B and 2B Fan Switch in Incorrect Position, (April 4, 2014)
performance in controlling worker access to radiologically significant areas and
* PER 876825, Vent Valve Found in Wrong Position, (April 25)
monitoring jobs in-progress during the week of the onsite inspection. The inspectors
* PER 877256, 'A' EGTS damper switch found in incorrect Position, (April 27)
also reviewed the procedural guidance for multi and extremity badging. For HRA tasks
* PER 878321, 'B' Train Purge Aligned with 'A' Train Radiation Monitor, (April 30)
involving significant dose rate gradients, the inspectors evaluated the use and placement
* PER 878588, Missing Locking Mechanism on Charging Valve (April 30)
of whole body and extremity dosimetry to monitor worker exposure. The inspectors
* PER 882745, Switch Error Alignment of Inverter during Testing, (May 9)
reviewed RWPs for use in airborne areas, ensuring the prescribed controls were
* PER 884002, Boric Acid Valve Found in Wrong Position, (May 13)
appropriate for the conditions as identified in radiological surveys and air samples. ED
* PER 884012, Danger-Tagged Switch Found in Wrong Position, (May 13)
alarm set points and worker stay times were evaluated against area radiation survey
* PER 885856, Incorrect Pressurizer Safety Valve Removed, (May 16)
results for containment and auxiliary building activities.
* PER 886066, Missed QC Hold Point, (May 17)
Risk Significant High Radiation Areas and Very High Radiation Area Controls: The
* PER 886765, RHR Valves Found in Wrong Position, (May 19)  The residents discussed this negative human performance trend with site management.
inspectors evaluated access barrier effectiveness for selected Locked High Radiation
Area (LHRA) and Very High Radiation Area (VHRA) locations. Changes to procedural
guidance for LHRA and VHRA controls were discussed with Radiation Protection (RP)
supervisors. During plant walk downs of the U2 Containment and Auxiliary Building, the
inspectors verified the posting/locking of LHRA/VHRA areas. Established radiological
controls (including airborne controls) were evaluated for selected tasks including work in
auxiliary building HRAs, and radiological waste processing and storage. In addition,
licensee controls for areas where dose rates could change significantly as a result of
plant shutdown and refueling operations were reviewed and discussed.
Radiation Worker Performance and RP Technician Proficiency: The inspectors
observed radiation worker performance through direct observation. Jobs observed
included routine waste packaging activities in the auxiliary building and routine survey
activities in the Auxiliary Building and Upper and Lower Containments in high radiation
and contaminated areas. The inspectors also observed health physics technicians
(HPTs) providing pre-job/RWP briefings, releasing material from the RCA, and providing
field coverage of jobs. Occupational workers adherence to selected RWPs and HPT
proficiency in providing job coverage were evaluated through direct observations and
interviews with licensee staff. ED alarm set points and worker stay times were evaluated
against area radiation survey results for reviewed RWPs.
Problem Identification and Resolution: PERs associated with radiological hazard
assessment and control were reviewed and assessed. The inspectors evaluated the
licensees ability to identify, characterize, prioritize, and resolve the identified issues in
accordance with procedure NPG-SPP-22-300, Corrective Action Program, (CAP)
Revision (Rev.) 1. The inspectors also evaluated the scope of the licensees internal
audit program and reviewed recent assessment results.
RP activities were evaluated against the requirements of Updated Final Safety Analysis
Report (UFSAR) Section 12; Technical Specifications (TS) Sections 6.12; 10 CFR Parts
19 and 20; and approved licensee procedures. Licensee programs for monitoring
materials and personnel released from the RCA were evaluated against 10 CFR Part 20
and IE Circular 81-07, Control of Radioactively Contaminated Material. Documents
reviewed are listed in the Attachment.
                                                                                      Enclosure


Most of the errors involved some form of procedural non-compliance. The licensee concurred with the observation and noted that they had also concurrently and independently (of the NRC resident staff) identified the same trend. This was documented in PER 884559 and generated on May 14. Immediate corrective actions to these errors included "stand-downs" emphasizing procedural compliance with the craft personnel and site-wide communications to remind staff to use "error reduction" tools when performing high risk activities. The inspectors noted that the licensee was aggressively dealing with these human performance deficiencies and a reasonable assurance exists that the trend can be reversed. Although these issues should be corrected, they constitute violations of minor significance that are not subject to enforcement action in accordance with Section 2 of the Enforcement Policy.
                                        22
b. Findings
  Introduction: The inspectors identified a Green, self-revealing, Non-cited Violation
  (NCV) of TS 6.12.1, High Radiation Area, for two examples of individuals entering an
  HRA without meeting the entry requirements as specified in TS 6.12.1.b and e.
  Description: On May 16, 2014, with the U2 reactor shutdown for refueling, a contract
  employee who was staging equipment and two contract decontamination technicians,
  working on separate jobs in the upper reactor containment, entered the same posted
  HRA near the reactor cavity. One of the decontamination technicians and the contractor
  staging equipment received dose rate alarms shortly after crossing the HRA boundary.
  Upon receiving the alarms, both individuals exited the area and contacted health physics
  (HP) as required. The two decontamination technicians were on RWP Number (No.)
  210061 with a dose setpoint of 31 mrem and dose rate setpoint of 91 milli-rem per hour
  (mrem/hr). The decontamination workers ED indicated a maximum dose rate of 97
  mrem/hr. The worker staging equipment was on RWP No. 240051 with a dose setpoint
  of 21 mrem and dose rate setpoint of 81 mrem/hr. That workers ED indicated a
  maximum dose rate of 133 mrem/hr. Accessible general area dose rates based on
  surveys in the area near the time of the event were as high as 160 mrem/hr at 30
  centimeters (cm).
  In both cases the workers had only been briefed for entry into Radiation Areas in the
  upper reactor containment and that expected dose rates in this area were 3-10 mrem/hr.
  They were not wearing the prescribed alarming dosimetry for an HRA entry, were not on
  an RWP that allowed HRA entry, and were not knowledgeable of the actual dose rates in
  the area. The licensee entered these events into their CAP (PERs 886668 and 886160).
  Immediate corrective actions included restricting worker access to the RCA and
  issuance of communications to the site and within the RP organization to reinforce roles
  in RWP adherence and access control.
  Analysis: The inspectors determined that entry into a HRA without meeting the entry
  requirements specified in TS 6.12.1 was a performance deficiency. This finding is more
  than minor because it is associated with the Occupational Radiation Safety Cornerstone
  attribute of Human Performance and adversely affects the cornerstone objective of
  ensuring adequate protection of worker health and safety from exposure to radiation
  from radioactive material during routine civilian nuclear reactor operation. Workers
  permitted entry into HRAs with inadequate knowledge of actual radiological conditions
  could receive unintended occupational exposures. The finding was evaluated using the
  Occupational Radiation Safety Significance Determination Process (SDP). The finding
  was not related to ALARA planning, nor did it involve an overexposure or substantial
  potential for overexposure, and the ability to assess dose was not compromised.
  Therefore, the inspectors determined the finding to be of very low safety significance
  (Green). The inspectors noted that the workers responded properly to the ED dose rate
  alarms thereby limiting their potential for unintended exposure. This finding involved the
  cross-cutting aspect of Human Performance, Avoid Complacency [H.12] because
  workers failed to apply appropriate error reduction tools while participating in pre-job
  briefs and prior to crossing the HRA boundaries.
                                                                                      Enclosure


{{a|4OA5}}
                                          23
==4OA5 ==
    Enforcement: TS 6.12.1, High Radiation Area, requires in part, entries into HRAs be
==OTHER ACTIVITIES==
    controlled by means of an RWP, associated radiation surveys, and other appropriate
    radiation protection equipment and measures and except for individuals qualified in RP
    procedures or escorted by such individuals, entry into such areas shall only be made
    after dose rates in the area have been determined and entry personnel are made
    knowledgeable of them. Contrary to the above, on May 16, 2014, workers entered a
    HRA using an RWP that did not allow HRA access, without using the proper alarming
    dosimetry, and without knowledge of the actual dose rates in the area. Because this
    violation was of very low safety significance and it was entered into the licensees CAP
    (PERs 886668 and 886160), this violation is being treated as an NCV, consistent with
    the Enforcement Policy: NCV 05000327/328, 2014003-02, Failure to Comply with Entry
    requirements to a HRA.
2RS2 Occupational ALARA Planning and Controls
a.  Inspection Scope
    Work Planning and Exposure Tracking: The inspectors reviewed work activities and
    their collective exposure estimates associated with the previous Unit 1 (U1) refueling
    outage, as well as the current U2 refueling outage 19 (U2R19). The U1 refueling outage
    19 (U1R19) and U2R19 ALARA planning packages (ALARA Plans) were reviewed for
    the following high collective exposure tasks: Refueling operations, Mechanical
    Maintenance, Plant Services, RP and Modifications. For the selected tasks, the
    inspectors reviewed the assumptions and basis for the dose rate and man-hour
    estimates. The inspectors discussed with ALARA staff the means by which wrench-
    hours were derived from the work order hours provided by craft supervision to ALARA
    staff. The inspectors verified the licensee had established several means to track and
    trend doses for ongoing work activities. The inspectors evaluated the incorporation of
    exposure reduction initiatives and operating experience, including historical post-job
    reviews, into RWP requirements. Collective dose data for selected tasks were
    compared with established dose estimates and evaluated against procedural criteria
    (trigger points) for additional ALARA review. Where applicable, changes to established
    estimates were discussed with ALARA planners and evaluated against work scope
    changes or unanticipated elevated dose rate. The inspectors discussed the operation of
    the Station ALARA Committee with the Site Vice President, the RP Manager and the
    ALARA Health Physicist. For ALARA Plans from U1R19, the inspectors compared the
    results achieved in terms of actual dose versus (vs.) planned dose and actual hours vs.
    estimated hours, reviewed in-progress and post-job ALARA reviews, and discussed the
    job planning, performance, and reviews with ALARA staff. For ALARA Plans associated
    with U2R19, the inspectors reviewed dose-to-date on select jobs, comparing estimates
    with actuals, and observed development of selected in-progress reviews.
    Source Term Reduction and Control: The inspectors reviewed the collective exposure
    three-year rolling average (TYRA) from 2011 - 2013 and reviewed historical outage
    collective exposure trends. Through interviews with licensee staff and document review,
    the inspectors assessed the licensees current activities related to source term reduction,
    including elevated zinc injection on U2, on-line chemistry using pH 7.4 to minimize
    corrosion product transport, extended reactor coolant pump run time to allow better
                                                                                      Enclosure


===.1 (Closed) Temporary Instruction 2515/182 - Review of the Industry Initiative to Control Degradation of Underground Piping and Tanks===
                                            24
    cleanup during shutdown, ultrasonic fuel cleaning, and response to fuel defects during
    previous operating cycles. The inspectors discussed the unexpectedly high activity of
    shutdown crud burst and changes expected in the short and long term relative
    abundances of Cobalt-58 and Cobalt-60 that would result from the change in the steam
    generator tube alloys and increasing the number of steam generator tubes by about a
    third. The dose implications of the various cobalt reduction activities coupled to the
    change in tube alloys for the next few outages was also discussed
    Radiation Worker Performance: Radiation worker performance was also observed and
    evaluated as part of Inspection Procedure 71124.01 and is documented in section 2RS1.
    While observing job tasks, the inspectors evaluated the use of remote technologies to
    reduce dose including teledosimetry and remote visual monitoring. Jobs observed were
    associated with the refueling and maintenance outage.
    Problem Identification & Resolution: Licensee CAP documents associated with ALARA
    planning and controls were reviewed and assessed. This included a review of selected
    Action Requests (PERs), self-assessments, and audits. The inspectors evaluated the
    licensees ability to identify, characterize, prioritize, and resolve the identified issues in
    accordance with procedure NPG-SPP-22.300, Corrective Action Program, Rev. 1. The
    inspectors also evaluated the scope and frequency of the licensees self-assessment
    program and reviewed recent assessment results.
    ALARA program activities were evaluated against the requirements of UFSAR Section
    12, Radiation Protection; TS Section 6.8, Procedures and Programs; 10 CFR Part 20;
    and approved licensee procedures. Documents reviewed are listed in the Attachment.
b.  Findings
    No findings were identified.
2RS3 In-Plant Airborne Radioactivity Control and Mitigation
a.  Inspection Scope
    Engineering Controls: The inspectors reviewed the use of temporary and permanent
    engineering controls to mitigate airborne radioactivity during U2R19 for steam generator
    (S/G) and U2 Thimble eddy current testing and the DAW Storage Building. The use of
    the U2 Containment Purge to minimize airborne concentrations in containment during
    refuel activities was discussed with licensee personnel. In addition, inspectors observed
    the placement and use of high efficiency particulate air negative pressure units, and air
    sampling equipment during observations of jobs in-progress.
    Use of Respiratory Protection Devices & Self-Contained Breathing Apparatus for
    Emergency Use: Inspectors reviewed the use of respiratory protection devices to limit
    the intake of radioactive material, including devices used for routine tasks and devices
    stored for use in emergency situations. Inspectors observed the physical condition of
    Self-Contained Breathing Apparatus (SCBA) units, negative pressure respirators
    (NPR)s, powered air purifying respirators and device components staged for routine and
                                                                                            Enclosure


====a. Inspection Scope====
                                          25
The inspectors conducted a review of records and procedures related to the licensee's program for buried piping and underground piping and tanks in accordance with Phase II of temporary instruction (TI) 2515/182 to confirm that the licensee's program contained attributes consistent with Sections 3.3.A and 3.3.B of Nuclear Energy Institute (NEI) 09-14, "Guideline for the Management of Buried Piping Integrity," Revision 3, and to confirm that these attributes were scheduled and/or completed by the NEI 09-14 deadlines. The inspectors interviewed licensee staff responsible for the buried piping program and reviewed program related activities to determine if the program attributes were accomplished in a manner which reflected acceptable practices in program management. The licensee's buried piping and underground piping and tanks program was inspected in accordance with paragraph 03.02.a of the TI and it was confirmed that activities, which correspond to completion dates specified in the program which have passed since the Phase 1 inspection was conducted, have been completed. The licensee's buried piping and underground piping and tanks program was inspected in accordance with paragraph 03.02.b of the TI and responses to specific questions found in http://www.nrc.gov/reactors/operating/ops-experience/buried-pipe-ti-phase-2-insp-req-2011-11-16.pdf were submitted to the NRC headquarters staff. Additionally, the inspectors reviewed the licensee's risk ranking process and implementation of the inspection plan using the guidance of paragraph 03.04 and 03.05 of the TI.
  emergency use throughout the plant. SCBA bottle air pressure, the number of units, and
  the number of spare masks and air bottles available was also evaluated by inspectors.
  The inspectors reviewed maintenance records for selected SCBA units for the past year
  and evaluated SCBA and NPR compliance with National Institute for Occupational
  Safety and Health certification requirements. The inspectors also reviewed records of
  Grade D (or better) air quality testing for supplied-air devices and SCBA bottles. In
  addition, the inspectors walked-down the compressor used for filling SCBA bottles. The
  inspectors reviewed the status and surveillance records of SCBAs staged for in-plant
  use during emergencies through review of records and walk-down of SCBA staged in
  the control room and selected locations.
  The inspectors verified the licensee had procedures in place to ensure that the use of
  respiratory protection equipment was ALARA when engineering controls were not
  practicable. Control room operators and fire brigade were interviewed on the use of the
  devices including SCBA bottle change-out and use of corrective lens inserts. Respirator
  qualification records were reviewed and cross checked for several control room
  operators. In addition, qualifications for individuals responsible for testing and repairing
  SCBA vital components were evaluated through review of training records. Selected
  maintenance records for SCBA units and air cylinder hydrostatic testing documentation
  were reviewed.
  The inspectors verified that the licensee has procedural requirements in place for
  evaluating air samples for the presence of alpha emitters and reviewed airborne
  radioactivity and contamination survey records for selected plant areas to ensure air
  samples are screened and evaluated per the procedure requirements.
  The inspectors walked-down the respirator issue and storage locations and verified that
  the equipment was appropriately stored and maintained. Records of monthly and
  quarterly inventory and inspection of the equipment were also reviewed by the
  inspectors. The inspectors discussed the process for issuing respirators, and verified
  that selected individuals qualified for respirator and/or SCBA use had completed the
  required training, fit-test, and medical evaluation.
  Problem Identification and Resolution: Licensee CAP documents associated with the
  control and mitigation of in-plant radioactivity were reviewed and assessed. This
  included review of selected PERs related to use of respiratory protection devices
  including SCBA. The inspectors evaluated the licensees ability to identify, characterize,
  prioritize, and resolve the identified issues in accordance with procedure NPG-SPP-22-
  300, Corrective Action Program, Rev.1. The inspectors also evaluated the scope of the
  licensees internal audit program and reviewed recent assessment results.
  RP activities were evaluated against the requirements UFSAR Section 12; 10 CFR Parts
  19 and 20; and approved licensee procedures. Documents and records reviewed are
  listed in the Attachment.
b. Findings
  No findings were identified.
                                                                                      Enclosure


====b. Findings====
                                          26
No findings were identified. Based upon the scope of the review described above, Phase 2 of TI-2515/182 was completed.
2RS4 Occupational Dose Assessment
a.  Inspection Scope
    External Dosimetry: The inspectors reviewed National Voluntary Laboratory
    Accreditation Program certification data and discussed program guidance for storage,
    processing, and evaluation of results for active and passive personnel dosimeters
    currently in use. Comparisons between ED and thermo-luminescent dosimeter data
    were discussed in detail. The inspectors reviewed ED alarm logs and reviewed
    licensees dosimeter incident reports and assessment actions for selected alarm events.
    Internal Dosimetry: Program guidance and assessment results for internally deposited
    radionuclides were reviewed. The inspectors reviewed selected Whole Body Count (in
    vivo) analyses from September 2012 to May 2014 as well as in-vitro assessments of
    tritium exposures to workers entering Unit 2 containment at power during this period.
    The licensees methods used in these assessments as well as the programs for
    collection and analysis of special bioassay samples were discussed with licensee staff.
    Special Dosimetric Situations: The inspectors evaluated the licensees use of multi-
    badging, extremity dosimetry, and dosimeter relocation within non-uniform dose rate
    fields and reviewed assessments for U2R19 for S/G maintenance workers. Worker
    monitoring in neutron areas was discussed with licensee staff. The inspectors also
    reviewed records of monitoring for declared pregnant workers from September 2012 to
    May 2014 and discussed monitoring guidance with dosimetry staff. In addition, methods
    for shallow dose assessments were reviewed and discussed.
    Problem Identification and Resolution: The inspectors reviewed and discussed selected
    CAP documents associated with occupational dose assessment. The inspectors
    evaluated the licensees ability to identify and resolve the issues in accordance with
    procedure NPG-SPP-22-300, Corrective Action Program, Rev.1. The inspectors also
    discussed the scope of the licensees internal audit program and reviewed recent
    assessment results.
    Occupational dose assessment activities were evaluated against the requirements of
    UFSAR Section 12; TS Section 6; 10 CFR Parts 19 and 20; and approved licensee
    procedures. Documents reviewed are listed in the Attachment.
b. Findings
    No findings were identified.
2RS5 Radiation Monitoring Instrumentation
a.  Inspection Scope
    Radiation Monitoring Instrumentation: During walk-downs of the auxiliary building and
    the RCA exit point, the inspectors observed installed radiation detection equipment.
    These included area radiation monitors (ARMs), liquid and gaseous effluent monitors,
                                                                                      Enclosure


{{a|4OA6}}
                                            27
==4OA6 Meetings, Including Exit==
      personnel contamination monitors (PCMs), small article monitors (SAMs), and portal
      monitors (PMs). The inspectors observed the physical location of the components and
      noted their material condition.
      In addition to equipment walk-downs, the inspectors reviewed source checks of various
      portable and fixed detection instruments, including ion chambers, teletectors, PCMs,
      SAMs, PMs, and an iSOLO alpha/beta counting system. The inspectors reviewed
      calibration records and evaluated alarm set-point values for PCMs, PMs, effluent
      monitors, an ARM, and a SAM. This included a sampling of instruments used for post-
      accident monitoring such as a containment high-range radiation monitor and effluent
      monitors for noble gas and iodine. The radioactive source used to calibrate an effluent
      monitor was evaluated for traceability to national standards. Calibration stickers on
      portable survey instruments were noted during inspection of the storage area for ready-
      to-use equipment. The most recent 10 CFR Part 61 analysis for DAW was reviewed to
      determine if calibration and check sources are representative of the plant source term.
      The inspectors also reviewed count room calibration records for a gamma spectroscopy
      germanium detector and a liquid scintillation detector.
      Effectiveness and reliability of selected radiation detection instruments were reviewed
      against details documented in the following: 10 CFR Part 20; NUREG-0737,
      Clarification of TMI Action Plan Requirements; UFSAR Chapters 11 and 12; and
      applicable licensee procedures.
      Problem Identification and Resolution: The inspectors reviewed selected PER reports in
      the area of radiological instrumentation. The inspectors evaluated the licensees ability
      to identify and resolve the issues in accordance with procedure NPG-SPP-22.300,
      Corrective Action Program, Rev. 1. Documents and records reviewed are listed in the
      Attachment.
  b. Findings
      No findings were identified.
4.    OTHER ACTIVITIES
      Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency
      Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Security
4OA1 Performance Indicator (PI) Verification
  a. Inspection Scope
      The inspectors sampled licensee submittals for the five PIs listed below for the period
      from January 2013 through March 2014 for both Unit 1 and Unit 2. Definitions and
      guidance contained in Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment
      Indicator Guideline, Revision 6, were used to determine the reporting basis for each data
      element in order to verify the accuracy of the PI data reported during that period.
                                                                                      Enclosure


=====Exit Meeting Summary=====
                                        28
  Cornerstone: Initiating Events
  *  Unplanned Scrams per 7000 Critical Hours
  *  Unplanned Scrams with Complications
  *  Unplanned Power Changes per 7000 Critical Hours
  The inspectors reviewed selected Licensee Event Reports (LERs) and portions of
  operator logs to verify whether the licensee had accurately identified the number of
  scrams and unplanned power changes that occurred during the previous four quarters
  for both units. The inspectors also reviewed the accuracy of the number of critical hours
  reported and the licensees basis for addressing the criteria for complications for each of
  the reported scrams. Documents reviewed are listed in the Attachment.
  Cornerstone: Occupational Radiation
  *  Occupational Exposure Control Effectiveness
  The inspectors reviewed PI data collected from November 2013 through May 2014, for
  the Occupational Exposure Control Effectiveness PI. For the reviewed period, the
  inspectors assessed PER records to determine whether HRA, VHRA or unplanned
  exposures, resulting in TS or 10 CFR 20 non-conformances, had occurred during the
  review period. The inspectors reviewed RCA exit transactions with exposures in excess
  of 100 milli-rem in order to determine compliance with the requirements of the RWP.
  The reviewed data were assessed against guidance contained in Nuclear Energy
  Institute (NEI) 99-02, "Regulatory Assessment Indicator Guideline," Rev. 6.
  Cornerstone: Public Radiation Safety
  *  Radiological Control Effluent Release Occurrences
  The inspectors reviewed the Radiological Control Effluent Release Occurrences PI
  results for the Public Radiation Safety Cornerstone from November 2013, through May
  2014. For the assessment period, the inspectors reviewed cumulative and projected
  doses to the public and PER documents related to Radiological Effluent Technical
  Specifications/Offsite Dose Calculation Manual issues.
b. Findings
  No findings were identified.
                                                                                    Enclosure


On July 9, 2014, the resident inspectors presented the inspection results to                 Mr. Simmons and other members of his staff, who acknowledged the findings. No proprietary information was discussed.
                                              29
4OA2 Problem Identification and Resolution
.1    Daily Review
  a. Inspection Scope
      As required by Inspection Procedure 71152, Identification and Resolution of Problems,
      and in order to help identify repetitive equipment failures or specific human performance
      issues for follow-up, the inspectors performed a daily screening of items entered into the
      licensees CAP. This was accomplished by reviewing the description of each new PER
      and attending daily management review committee meetings.
  b. Findings and Observations
      No findings were identified.
.2    Selected Issue Follow-up: Pressurizer Thermal Limit Exceeded
  a. Inspection Scope
      The inspectors performed an in-depth review of PER 809100. On November 9, 2013,
      during a plant heat-up on Unit 1, the pressurizer thermal limits were exceeded on two
      separate occurrences, contrary to the Technical Requirements Manual (TRM) Section
      3.9.2. Westinghouse performed an analysis of the event and determined that the
      existing pressurizer design basis analysis bounded this event and there was no increase
      in the limiting primary stress or the primary-plus-secondary stress range from this event.
      As a result of this event, the licensee developed an Apparent Cause Analysis (ACE).
      The inspectors reviewed the actions taken to determine if the licensee had adequately
      addressed the following attributes.
      *    Complete, accurate and timely identification of the problem
      *    Evaluation and disposition of operability and reportability issues
      *    Consideration of previous failures, extent of condition, generic or common cause
          implications
      *    Prioritization and resolution of the issue commensurate with safety significance
      *    Identification of the root cause and contributing causes of the problem
      *    Identification and implementation of corrective actions commensurate with the safety
          significance of the issue
  b. Findings
      There were no findings identified during this review. The inspectors determined that the
      ACE was probing and involved an extent of condition review, a safety culture analysis,
      and operating experience review. The ACE also brought to light that the crew that
      performed the pressurizer heat up did not realize the limit violation. This fact was
      actually noted by the night shift crew when reviewing data. The ACE also analyzed a
      near miss that occurred on November 13, 2013. In this instance, the plant was being
                                                                                        Enclosure


ATTACHMENT: 
                                          30
      returned to cold iron conditions due to a pressurizer power operated relief failure and the
      operators were challenged in maintaining cool-down limits of the pressurizer. Although
      no limits were exceeded the November 13 incident, the ACE noted several weaknesses
      in the evolution. The ACE ultimately led to the development of several corrective
      actions; including procedural changes designed strengthen the operators awareness of
      pressurizer pressure control, development of a vendor-performed stress analysis of the
      event, and addition of the event to the operations training program in order to share
      information with other crews.
.3    Semi-Annual Trend Review
  a. Inspection Scope
      As required by Inspection Procedure 71152, the inspectors performed a semi-annual
      review of the licensees corrective action program and associated documents to identify
      trends that could indicate the existence of a more significant safety issue. The
      inspectors review was focused on repetitive equipment issues, but also included
      licensee trending efforts and licensee human performance results. The inspectors
      review nominally considered the twelve-month period of July 2013 through June 2014,
      although some examples expanded beyond those dates when the scope of the trend
      warranted. Specifically, the inspectors considered the results of daily inspector
      screening discussed in Section 4OA2.1 and reviewed licensee trend reports for the
      period in order to determine the existence of any adverse trends that the licensee may
      not have previously identified. This activity constituted one inspection sample for Semi-
      annual Trend Review.
  b. Findings and Observations
      No findings were identified. The inspectors noted a negative trend regarding human
      performance errors. During the daily reviews, the inspectors noted an increase in
      human error events. The inspectors then performed a more detailed review of the trend
      under the semi-annual trend review required by IP 71152. The inspectors concluded
      there were at least eleven of these events that occurred in the last three months. The
      inspectors noted this was more than the typical amount of error-related incidents
      observed during a quarter. The below abbreviated list of PERs involved several human
      performance related and mis-positioning events as well as procedural non-compliance.
      *    PER 868301, EDG 1B and 2B Fan Switch in Incorrect Position, (April 4, 2014)
      *    PER 876825, Vent Valve Found in Wrong Position, (April 25)
      *    PER 877256, A EGTS damper switch found in incorrect Position, (April 27)
      *    PER 878321, B Train Purge Aligned with A Train Radiation Monitor, (April 30)
      *    PER 878588, Missing Locking Mechanism on Charging Valve (April 30)
      *    PER 882745, Switch Error Alignment of Inverter during Testing, (May 9)
      *    PER 884002, Boric Acid Valve Found in Wrong Position, (May 13)
      *    PER 884012, Danger-Tagged Switch Found in Wrong Position, (May 13)
      *    PER 885856, Incorrect Pressurizer Safety Valve Removed, (May 16)
                                                                                        Enclosure


=SUPPLEMENTAL INFORMATION=
                                            31
      * PER 886066, Missed QC Hold Point, (May 17)
      * PER 886765, RHR Valves Found in Wrong Position, (May 19)
      The residents discussed this negative human performance trend with site management.
      Most of the errors involved some form of procedural non-compliance. The licensee
      concurred with the observation and noted that they had also concurrently and
      independently (of the NRC resident staff) identified the same trend. This was
      documented in PER 884559 and generated on May 14. Immediate corrective actions to
      these errors included stand-downs emphasizing procedural compliance with the craft
      personnel and site-wide communications to remind staff to use error reduction tools
      when performing high risk activities. The inspectors noted that the licensee was
      aggressively dealing with these human performance deficiencies and a reasonable
      assurance exists that the trend can be reversed. Although these issues should be
      corrected, they constitute violations of minor significance that are not subject to
      enforcement action in accordance with Section 2 of the Enforcement Policy.
4OA5 OTHER ACTIVITIES
.1    (Closed) Temporary Instruction 2515/182 - Review of the Industry Initiative to Control
      Degradation of Underground Piping and Tanks
  a. Inspection Scope
      The inspectors conducted a review of records and procedures related to the licensees
      program for buried piping and underground piping and tanks in accordance with
      Phase II of temporary instruction (TI) 2515/182 to confirm that the licensees program
      contained attributes consistent with Sections 3.3.A and 3.3.B of Nuclear Energy
      Institute (NEI) 09-14, Guideline for the Management of Buried Piping Integrity,
      Revision 3, and to confirm that these attributes were scheduled and/or completed by
      the NEI 09-14 deadlines. The inspectors interviewed licensee staff responsible for the
      buried piping program and reviewed program related activities to determine if the
      program attributes were accomplished in a manner which reflected acceptable
      practices in program management.
      The licensees buried piping and underground piping and tanks program was inspected
      in accordance with paragraph 03.02.a of the TI and it was confirmed that activities,
      which correspond to completion dates specified in the program which have passed
      since the Phase 1 inspection was conducted, have been completed. The licensees
      buried piping and underground piping and tanks program was inspected in accordance
      with paragraph 03.02.b of the TI and responses to specific questions found in
      http://www.nrc.gov/reactors/operating/ops-experience/buried-pipe-ti-phase-2-insp-req-
      2011-11-16.pdf were submitted to the NRC headquarters staff. Additionally, the
      inspectors reviewed the licensees risk ranking process and implementation of the
      inspection plan using the guidance of paragraph 03.04 and 03.05 of the TI.
                                                                                        Enclosure


==KEY POINTS OF CONTACT==
                                        32
b.  Findings
    No findings were identified. Based upon the scope of the review described above,
    Phase 2 of TI-2515/182 was completed.
4OA6 Meetings, Including Exit
    Exit Meeting Summary
    On July 9, 2014, the resident inspectors presented the inspection results to
    Mr. Simmons and other members of his staff, who acknowledged the findings. No
    proprietary information was discussed.
    ATTACHMENT: SUPPLEMENTAL INFORMATION
                                                                                  Enclosure


===Licensee personnel===
                              SUPPLEMENTAL INFORMATION
: [[contact::J. Carlin]], Site Vice President  
                                KEY POINTS OF CONTACT
: [[contact::A. Day]], Chemistry Manager  
Licensee personnel
: [[contact::D. Erb]], Work Control Manager  
J. Carlin, Site Vice President
: [[contact::B. Gann]], Dosimetry/Instruments Supervisor  
A. Day, Chemistry Manager
: [[contact::M. Henderson]], ISI Program Engineer  
D. Erb, Work Control Manager
: [[contact::J. Johnson]], Program Manager Licensing  
B. Gann, Dosimetry/Instruments Supervisor
: [[contact::T. Johnston]], Radiation Protection Support Manager  
M. Henderson, ISI Program Engineer
: [[contact::K. Loomis]], Site Program Owner  
J. Johnson, Program Manager Licensing
: [[contact::T. Marshall]], Operations Manager  
T. Johnston, Radiation Protection Support Manager
: [[contact::M. McBrearty]], Licensing Manager  
K. Loomis, Site Program Owner
: [[contact::T. Noe]], Director Safety and Licensing  
T. Marshall, Operations Manager
: [[contact::W. Pierce]], Site Engineering Director  
M. McBrearty, Licensing Manager
: [[contact::P. Pratt]], Maintenance Manager  
T. Noe, Director Safety and Licensing
: [[contact::R. Rice]], Radiation Protection Manager  
W. Pierce, Site Engineering Director
: [[contact::J. Rolph]], Radiation Protection Technical Support Superintendent  
P. Pratt, Maintenance Manager
: [[contact::P. Simmons]], Plant Manager  
R. Rice, Radiation Protection Manager
: [[contact::K. Smith]], Director of Training  
J. Rolph, Radiation Protection Technical Support Superintendent
: [[contact::C. Summers]], Health Physicist-ALARA  
P. Simmons, Plant Manager
K. Smith, Director of Training
C. Summers, Health Physicist-ALARA
NRC personnel
S. Lingam, Project Manager, Office of Nuclear Reactor Regulation
                    LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
05000327,328/2014003-01              NCV          Failure to Perform Visual Examination of the
                                                  Unit 1 and Unit 2 CRDM Seismic Plate
                                                  Supports (Section 1R08)
05000327,328/2014003-02              NCV          Failure to Comply with Entry requirements
                                                  to a HRA. (Section 2RS1)
Closed
2515/182                            TI            Review of the Industry Initiative to Control
                                                  Degradation of Underground Piping and
                                                  Tanks, Phase II (Section 4OA5.1)
                                                                                      Attachment


===NRC personnel===
                            LIST OF DOCUMENTS REVIEWED
: [[contact::S. Lingam]], Project Manager, Office of Nuclear Reactor Regulation
Section R04: Equipment Alignment
Procedures
0-GO-16, System Operability Checks, Rev 19
Section R05: Fire Protection
Procedures
SQN-FPR-Part-II, SQN Fire Protection Report Part II - Fire Protection Plan, Revision 28
Other documents
TUR-0-706-01, Fire Protection Pre-Fire Plans Turbine Building - El. 706, Revision 3
TUR-0-706-02, Fire Protection Pre-Fire Plans Turbine Building - El. 706, Revision 3
CON-0-706-00, Fire Protection Pre-Fire Plans Control Building - El. 706, Revision 6
CON-0-706-00, Fire Protection Pre-Fire Plans Control Building - El. 732, Revision 7
Section R06: Flood Protection Measures
Other documents
TVA letter to NRC dated May 4, 2007. TVA response to GL 2007-01
48N1225
47W852-1
SQN-SQS4-0056, Moderate Energy Line Break Flooding Study, Revision 2
Section R08: Inservice Inspection
Drawings
CCD No. 2-2H63-0006-01, Residual Heat Removal System Pipe Support, Rev. 0
CCD No. 2-2-H2O-0020-01, Sequoyah Nuclear Plant, Safety Injection System Pipe Supports,
R-0
1095E46, Sequoyah Nuclear Plant No. 2, CRDM Seismic Support Platform General Assembly,
Sheet 1 of 2, Rev. 6
1, 2-47W813-1-ISI, Flow Diagram Reactor Coolant System, Rev. 7
2-47K406-112, N2-62-12A Isometric, Static, Thermal, and Seismic Analysis of CVCS Piping,
Rev.1
ISI-0401-C-02, Sequoyah Nuclear Plant Unit-2 Steam Generator Replacement, Rev. 3
48N427, Structural Steel Equipment Supports Upper Steam Generator Support, Rev. 15
48N431, Structural Steel Equipment Supports Upper Steam Generator Support Details, Rev. 20
DCA No. D22479-3001, Drawing Change Authorization, DCN D22479A, Page 85, Rev. 2
DCA No. D22479-3002, Drawing Change Authorization, DCN D22479A, Page 86, Rev. 1
DCA No. D22479-3003, Drawing Change Authorization, DCN D22479A, Page 87, Rev. 2
DCA No. D22479-3005, Drawing Change Authorization, DCN D22479A, Page 89, Rev. 2
CCD No. 2-2-H47-0104-01, Steam Generator Blowdown System Pipe Supports, Rev. 0
Procedures
0-PI-SLT-068-200.0, Reactor Building Post Shutdown Leakage Examination, Rev. 4
2-PVC-068-0340B, Preventative Maintenance Work Instruction, PM011442000, Attachment A
used by WO 114734912, dated 5/12/14
0-MI-MRR-068-006.0, Sequoyah Nuclear Plant, Installation of Reactor Pressure Vessel Head
and Attachments, Appendix G Seismic Tie Data Sheet, Rev. 38
                                                                                    Attachment


==LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED==
                                          3
2-SI-SXI-068-201.0, Leakage Test of the Reactor Coolant Pressure Boundary, 1/21/2014
N-PT-9, Liquid Penetrant of ASME and ANSI Code Components and Welds, Rev. 0037
N-VT-1, Visual Examination Procedure for ASME Section XI Preservice and Inservice, Rev. 5
Areva 03-9052292, Operating Instructions for RANGER in recirculating Steam Generator,
Rev. 15
Areva 03-9187284, Utilizing a Personal Computer Platform for Eddy Current Acquisition Data
Functions, Rev. 1
NEDP-16, Steam Generator Program, Rev. 15
0-MI-MXX-068-005.0, Steam Generator Primary Side Maintenance Activities, Rev. 24
0-MI-MXX-003-002.0, Steam Generator Secondary Side Maintenance Activities, Rev. 15
0-SI-SXI-068-114.3, Steam Generator Tubing Inservice Inspection and Augmented Inspections,
Rev. 1
2-SI-CEM-068-137.5, Primary to Secondary Leakage via Steam Generators, Unit 2, Rev. 13
Engineering/Technical Evaluations
PER 888330, Boric Acid Leakage Evaluation, Reactor Cavity Nozzle Cover Seal leaking into
Keyway, 6/24/14
PER 890230, Evaluation of Boric Acid Corrosion Damage, 2-SNUB-068-RCH30, 6/7/14
Sequoyah PER 743110, Degraded Unit 2, Ice Condenser Due To Recurring Frost Accumulation
on Intermediate Deck Doors, Event Date, 5/08/13
SQN PER 889645, Equipment Apparent Cause Evaluation for Compression Fitting Leak, Event
Date, 6/24/14
NOI-2-SQ-432, Available Margins in Pipe Support Attributes, 5/26/14
Corrective Action Documents
PER 888991, Observation made during NRC ISI - Boric Acid Inspection, 5/28/14
PER 889400, Determine whether CRDM Seismic Support should be examined under Section
XI, 5/23/14
PER 899941, Failure to Quarantine Failed Part for Analysis, 6/17/2014
PER 743110, Degraded Non-conforming condition for PDO on RCS leakage and Ice on
Intermediate deck doors (IDD), 7/19/13
SR888431, Loose Hydraulic lines on Snubbers, 5/22/2014
PER 487507, SQN review/Westinghouse NSAL-12-1 SG Channel Head Degradation, 2012
PER 889451, Discoloration in Steam Generator Primary Bowls, 2014
SR 890656, Steam Generator Secondary Side Inspection and Sludge Lancing
SR 891631, EPRI ETSS not referred in site ETSS, 2014
SR 891633, Steam Generator ECT Secondary Analyst did not call wear out and proximity
indications, 2014SR 900540, Evaluate SEQ Primary to Secondary Leakrate Detection Limits,
2014
Other Documents
Penetration Number 56, RPV Head Penetration UT Data Sheet, 12/7/06
Penetration Number 56, RPV Head Penetration UT Data Sheet, 05/18/14
Penetration Number 60, RPV Head Penetration UT Data Sheet, 05/18/14
Penetration Number 53, RPV Head Penetration UT Data Sheet, 12/06/06
Penetration Number 53, RPV Head Penetration UT Data Sheet, 05/19/14
R-6069, TVA Record of Liquid Penetrant Examination, 2SIH-020-IA, 4/30/99
                                                                                  Attachment


===Opened and Closed===
                                            4
: 05000327,328/2014003-01 NCV Failure to Perform Visual Examination of the Unit 1 and Unit 2 CRDM Seismic Plate Supports (Section 1R08) 
R0114, TVA Liquid Penetrant Examination, Reinspection Summary No. 01961-ISI-SQN,
: 05000327,328/2014003-02  NCV  Failure to Comply with Entry requirements
for 2-SIH-020-IA, 5/19/14
to a HRA. (Section 2RS1)
R0105, TVA Liquid Penetrant Examination, Inspection Summary No. 01934-ISI-SQN2,
or 2-CVCH-006-IA, 5/18/14
NPG-SPP-09.1, ASME Code and Augmented Programs, Attachment 8, Form NPG-SPP-09.1-2,
for Component ID, 2-SIH-020-IA, 5/19/14
NPG-SPP-09.1, ASME Code and Augmented Programs, Attachment 8, Form NPG-SPP-09.1-2,
for Component ID, 2-CVCH-585, 5/12/14
System 068, Reactor Coolant System Health Report, 2/1/2014 - 5/31/2014
0-SI-DXI-000-114.3, Attachment 5, Unit-2 Examination Schedule for ASME Class 1, 2, 3
Components, 5/9/14
TVA Report No. R0105, Summary No. 01934-ISI-SQN2, Liquid Penetrant Examination
Summary for Component ID 2-CVCH-006-IA, Category B-K/B10.20, Integral Attachment,
5/18/14
Work Order No. 112354373, Valve SQN-2-VLV-001-0817 Replacement, 3/25/13
R0041, TVA Record of Visual Examination, 2-CVCH-585, 5/6/2014
R0086, TVA Record of Visual Examination, 2-CVCH-584, 5/15/2014
R0094, TVA Record of Visual Examination, 2-CVCH-586, 5/14/2014
R0151, Ultrasonic Piping Examination Data Sheet, FDF-011A, 5/24/2014
R0152, Ultrasonic Piping Examination Data Sheet, FDF-010C, 5/24/2014
R0170, TVA Record of Visual Examination, SGH-4-1, 5/28/14
Candidate No. 3237861, EPRI Performance Demonstration Initiative Program Qualifications,
1/14/11
MWK7861, IHI Southwest Technologies Inc. Certificate of Qualification, 2/22/2013
H14132981, Certificate of Calibration, M&TE ID No. E41820, 4/14/2013
VT-1, Certificate of Method Qualification Record for BMNO6QGPV, Expires, 11/28/2014
VT-3, Certificate of Method Qualification Record for BMNO6QGPV, Expires, 11/28/2014
VT-3, Certificate of Method Qualification Record for D880WSO0D, Expires, 10/5/2014
0-SI-MFT-000-001.0, Appendix E Page 1, Snubber Functional Testing, SQN-2-SNUB-015-
SGBH104, 5/18/14
Report No. SCV-0001, Visual Examination of IWE Interfaces, Moisture Barrier, 5/20/2014
Report No. SCV-0004, Visual Examination of IWE Interfaces, Moisture Barrier, 5/5/2014
Report No. SCV-0005, Visual Examination of IWE Interfaces, Moisture Barrier, 5/5/2014
0-TI-DXX-000-097.1, Boric Acid Corrosion Control Program, Rev. 0009
NPG-SPP-09.7.4, Boric Acid Corrosion Control Program, Rev. 0001
0-TI-SPT-000-301.0, ASME Section XI Pressure testing Program Basis Document, Rev. 0004
0-TI-RVI-000-301.0, Sequoyah Unit 1 & 2, PWR Reactor Vessel Internals Inspection Program,
Rev. 0
Sequoyah Unit 2 Control Room, Total Unidentified Leakage Logs, 12/28/2012 thru 5/10/2014
Areva Use of Appendix H and Appendix I Qualified Techniques Sequoyah U2R19 Refueling
Outage, Rev. 0, May 2014
Areva 51-91988290-00, Sequoyah U2R19 Steam Generator Degradation Assessment,
Rev. 0, May 2014
Areva SQN 2C19 Analyst Training Instructions, Rev. 0
Areva 54-ISI-400-021, Eddy Current Inspection Multi-Frequency Eddy Current Examination of
Tubing, June 2013
Areva 51-9221442-000, Sequoyah Unit 2 EOC19 SG ECT Inspection Plan
Areva ETSS_BOB1, Areva Examination Technical Specification Sheet for Bobbin Probe, Rev. 0
                                                                                  Attachment


===Closed===
                                            5
: 2515/182    TI  Review of the Industry Initiative to Control        Degradation of Underground Piping and
Areva ETSS_RPC1, Areva Examination Technical Specification Sheet for MRPC Probe, Rev. 0
: Tanks, Phase II (Section 4OA5.1) 
Areva ETSS_RPC2, Areva Examination Technical Specification Sheet for MRPC Probe, Rev. 0
Areva ETSS_Array1, Areva Examination Technical Specification Sheet for MRPC Probe, Rev. 0
B85 130429 005, Sequoyah Unit 2 Refueling 18 Operational Assessment, Rev. 1
Calibration Records for Eddy Current Tester Miz80i Serial Numbers: 39, 21, 71, 36, 73, 91
Certificate of Conformance for Eddy Current Probes, Serial Numbers 653790, 652262, 655350,
653784, 652251, 653863, 652265, 652242
Calibration Standard for ASME 21095, 21099, 21100, 21096, EDM 9173936, 21086,
ARRAY 9173939
Personnel Qualification Records for Qualified Data Analysts: W. Bridforth, D. Cornell,
N. Farenbaugh, J. Janet Sr, R. Lee, G. Manley, W. McMillan, S. Merriam, E. Miranda,
R. Miranda, J. Parrish, J. Oliver, A. Richardson, T. Shulter, J. Sordini, L. Tobin, D. Torres
Personnel Qualification Records of TVA Steam Generator Program Personnel: J. Mayo,
W. James
SQN-ENG-F-10-02, Self-Assessment on Steam Generator Program, April 2010
SQN-ENG-S-11-91, Benchmarking Report on U2R17 NRC Inservice Inspection Readiness,
March 2011
SQN-CEM-S-10-015, Self-Assessment on EPRI Secondary Water Chemistry Guidelines,
July 2010
Sequoyah Nuclear Plant Unit 2, Replacement Steam Generator Eddy Current Examination
Guideline, Rev. 1
Structural Integrity Associates, Report No. 1400660.401.R0, Independent Review of
Westinghouse LTR-SGMMP-14-27, Assessment of Discolorations on Replacement Steam
Generator Channel Head Cladding at Sequoyah Unit 2, dated May 30, 2014
Section R12: Maintenance Effectiveness
Procedures
TI-4, Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting -
10CFR50.65, Revision 25
Section R13: Maintenance Risk Assessments and Emergent Work Evaluation
Procedures
NPG-SPP-07.0, Work Management, Revision 0
NPG-SPP-07.1, On Line Work Management, Revision 9
NPG-SPP-07.3, Work Activity Risk Management Process, Revision 13
NPG-SPP-07.11.1, Equipment Out of Service Management, Revision 9
Section R15: Operability Evaluations
Procedures
NEDP-22, Functional Evaluations, Rev. 15
OPDP-8, Limiting Conditions for Operation Tracking, Rev. 16
NPG-SPP-03.5, Regulatory Reporting Requirements, Rev. 10
Section R19: Post Maintenance Testing
Procedures
MMDP-1, Maintenance Management System, Rev. 20
MMDP-3, Guidelines for Planning and Execution of Troubleshooting Activities, Rev. 6
NPG-SPP-6.5, Foreign Material Control, Rev. 4
                                                                                        Attachment


==LIST OF DOCUMENTS REVIEWED==
                                          6
Section R04:
NPG-SPP-6.1, Work Order Process Initiation, Rev. 2
: Equipment Alignment Procedures 0-GO-16, System Operability Checks, Rev 19
NPG-SPP-06.3, Pre-/Post-Maintenance Testing, Rev. 1
: Section R05:
NPG-SPP-06.9, Testing Programs, Rev. 0
: Fire Protection Procedures
NPG-SPP-06.9.1, Conduct of Testing, Rev. 8
: SQN-FPR-Part-II, SQN Fire Protection Report Part II - Fire Protection Plan, Revision 28 
NPG-SPP-06.9.3, Post-Modification Testing, Rev. 5
: Other documents
Section 2RS1: Radiological Hazard Assessment and Exposure Controls
: TUR-0-706-01, Fire Protection Pre-Fire Plans Turbine Building - El. 706, Revision 3
Procedures, Guidance Documents, and Manuals
: TUR-0-706-02, Fire Protection Pre-Fire Plans Turbine Building - El. 706, Revision 3
NPG-SPP-05.1 Radiological Controls Revision (Rev.) 003
: CON-0-706-00, Fire Protection Pre-Fire Plans Control Building - El. 706, Revision 6
NPG-SPP-05.1.1, Alpha Radiation Monitoring Program, Rev. 003
: CON-0-706-00, Fire Protection Pre-Fire Plans Control Building - El. 732, Revision 7
NPG-SPP-05.6, Controlling Byproduct and Source Material, Rev. 002
: Section R06:
O-SI-RCI-000-056.0, Byproduct Material Inventory and Sealed Source Leak Test Rev. 016
: Flood Protection Measures Other documents TVA letter to NRC dated May 4, 2007. TVA response to
RCI-14, Radiation Work Permit (RWP) Program, Rev. 058
: GL 2007-01 48N1225
RCI-15, Radiological Postings Rev. 026
: 47W852-1
RCI-21, Control of Radioactive Materials, Rev. 019
: SQN-SQS4-0056, Moderate Energy Line Break Flooding Study, Revision 2
RCI-22, Contamination Control Rev. 024
: Section R08:
RCI-24, Control of Very High Radiation Areas Rev. 014
: Inservice Inspection Drawings CCD No. 2-2H63-0006-01, Residual Heat Removal System Pipe Support, Rev. 0
RCI-28, Control of Locked High Radiation Areas Rev. 015
: CCD No. 2-2-H2O-0020-01, Sequoyah Nuclear Plant, Safety Injection System Pipe Supports,  
RCI-29, Control of Radiation Protection Keys, Rev. 016
: R-0
RCI-201, Radiation and Contamination Surveys, Rev. 015
: 1095E46, Sequoyah Nuclear Plant No. 2, CRDM Seismic Support Platform General Assembly, Sheet 1 of 2, Rev. 6 1, 2-47W813-1-ISI, Flow Diagram Reactor Coolant System, Rev. 7 
RCI-202, Airborne Radioactivity Surveys Rev. 008
: 2-47K406-112, N2-62-12A Isometric, Static, Thermal, and Seismic Analysis of CVCS Piping, Rev.1
RCI-204, Radiological Surveys of Equipment and Materials Leaving the RCA, Rev. 008
: ISI-0401-C-02, Sequoyah Nuclear Plant Unit-2 Steam Generator Replacement, Rev.
RCI-404, Radiation Protection Requirements for Remote Job Coverage, Rev. 001
: 3
RCI-417, Radiological Monitoring of the Hydrogen Peroxide Injection Crud Burst, Rev. 001
: 48N427, Structural Steel Equipment Supports Upper Steam Generator Support, Rev. 15 48N431, Structural Steel Equipment Supports Upper Steam Generator Support Details, Rev. 20 DCA No. D22479-3001, Drawing Change Authorization, DCN D22479A, Page 85, Rev. 2
RCDP-1, Conduct or Radiological Control Rev. 005
: DCA No. D22479-3002, Drawing Change Authorization, DCN D22479A, Page 86, Rev. 1
Records and Data
: DCA No. D22479-3003, Drawing Change Authorization, DCN D22479A, Page 87, Rev. 2
0-SE-RCI-000-056.0 Byproduct Material Inventory and Sealed Source Leak Test, 11/23/2013
: DCA No. D22479-3005, Drawing Change Authorization, DCN D22479A, Page 89, Rev. 2 CCD No. 2-2-H47-0104-01, Steam Generator Blowdown System Pipe Supports, Rev. 0
0-SE-RCI-000-056.0 Byproduct Material Inventory and Sealed Source Leak Test, 05/02/2014
===Procedures===
0-TI-NUC-000-002.0, Storing Material in Spent Fuel Pool or New Fuel Vault, Rev. 0022,
: 0-PI-SLT-068-200.0, Reactor Building Post Shutdown Leakage Examination, Rev. 4
Attachment-1, Inventory of Materials in Spent Fuel Pool, Cask Loading Area, and New Fuel
: 2-PVC-068-0340B, Preventative Maintenance Work Instruction, PM011442000, Attachment "A" used by
Vault, dated 02/06/2014.
: WO 114734912, dated 5/12/14
2013 Sequoyah Radiation Protection Annual Report conducted per NPG-SPP-05.1,
: 0-MI-MRR-068-006.0, Sequoyah Nuclear Plant, Installation of Reactor Pressure Vessel Head and Attachments, Appendix "G" Seismic Tie Data Sheet, Rev. 38 
Radiological Controls, Section 3.2, Program Monitoring Evaluation and Oversight
: Attachment 2-SI-SXI-068-201.0, Leakage Test of the Reactor Coolant Pressure Boundary, 1/21/2014 N-PT-9, Liquid Penetrant of ASME and ANSI Code Components and Welds, Rev. 0037 N-VT-1, Visual Examination Procedure for ASME Section XI Preservice and Inservice, Rev. 5
Airborne Radiation Survey (ARS) Number (No.) 051514008, U-2 Lower Seal Table During Eddy
: Areva 03-9052292, Operating Instructions for RANGER in recirculating Steam Generator,
Current Test, dated 05/15/2014
: Rev. 15
ARS No. 051514004, U-2 Lower Containment Raceway Routine, dated 05/15/2014
: Areva 03-9187284, Utilizing a Personal Computer Platform for Eddy Current Acquisition Data Functions, Rev. 1
ARS No. 051514002, U-2 Lower Containment Routine @Elevation 679, dated 05/15/2014
: NEDP-16, Steam Generator Program, Rev. 15 0-MI-MXX-068-005.0, Steam Generator Primary Side Maintenance Activities, Rev. 24
ARS No. 051514003, U-2 Lower Containment Routine @IPCW, dated 05/15/2014
: 0-MI-MXX-003-002.0, Steam Generator Secondary Side Maintenance Activities, Rev. 15
ARS No. 052114017, U-2 S/G#1 Laydown Area During Insert Removal, dated 05/21/2014
: 0-SI-SXI-068-114.3, Steam Generator Tubing Inservice Inspection and Augmented Inspections, Rev. 1 2-SI-CEM-068-137.5, Primary to Secondary Leakage via Steam Generators, Unit 2, Rev. 13 
ARS No. 052114018, U-2 S/G#1 During Insert Removal, dated 05/21/2014
: Engineering/Technical Evaluations
ARS No. 052114019, U-2 S/G#4 During Insert Removal, dated 05/21/2014
: PER 888330, Boric Acid Leakage Evaluation, Reactor Cavity Nozzle Cover Seal leaking into Keyway, 6/24/14
ARS No. 052114022, U-2 S/G#3 During Insert Removal, dated 05/21/2014
: PER 890230, Evaluation of Boric Acid Corrosion Damage, 2-SNUB-068-RCH30, 6/7/14 Sequoyah
ARS No. 052114023, U-2 S/G#2 During Insert Removal, dated 05/21/2014
: PER 743110, Degraded Unit 2, Ice Condenser Due To Recurring Frost Accumulation on Intermediate Deck Doors, Event Date, 5/08/13 
ARS No. 052114024, U-2 S/G#3 Laydown Area During Insert Removal, dated 05/21/2014
: SQN
ARS No. 052114025, U-2 S/G#1 Laydown Area Back-up Sample, dated 05/21/2014
: PER 889645, Equipment Apparent Cause Evaluation for Compression Fitting Leak, Event Date, 6/24/14
ARS No. 052114026, U-2 S/G#3 Laydown Area Back-up Sample, dated 05/21/2014
: NOI-2-SQ-432, Available Margins in Pipe Support Attributes, 5/26/14
ARS No. 052114027, U-2 S/G#3 Primary PlatformBack-up Sample, dated 05/21/2014
: Corrective Action Documents
                                                                                  Attachment
: PER 888991, Observation made during NRC ISI - Boric Acid Inspection, 5/28/14
: PER 889400, Determine whether CRDM Seismic Support should be examined under Section XI, 5/23/14
: PER 899941, Failure to Quarantine Failed Part for Analysis, 6/17/2014
: PER 743110, Degraded Non-conforming condition for PDO on RCS leakage and Ice on Intermediate deck doors (IDD), 7/19/13
: SR888431, Loose Hydraulic lines on Snubbers, 5/22/2014
: PER 487507, SQN review/Westinghouse
: NSAL-12-1 SG Channel Head Degradation, 2012
: PER 889451, Discoloration in Steam Generator Primary Bowls, 2014
: SR 890656, Steam Generator Secondary Side Inspection and Sludge Lancing
: SR 891631, EPRI ETSS not referred in site ETSS, 2014
: SR 891633, Steam Generator ECT Secondary Analyst did not call wear out and proximity indications, 2014SR
: 900540, Evaluate SEQ Primary to Secondary Leakrate Detection Limits,
: 2014
===Other Documents===
: Penetration Number 56, RPV Head Penetration UT Data Sheet, 12/7/06
: Penetration Number 56, RPV Head Penetration UT Data Sheet, 05/18/14
: Penetration Number 60, RPV Head Penetration UT Data Sheet, 05/18/14
: Penetration Number 53, RPV Head Penetration UT Data Sheet, 12/06/06 Penetration Number 53, RPV Head Penetration UT Data Sheet, 05/19/14 R-6069, TVA Record of Liquid Penetrant Examination, 2SIH-020-IA, 4/30/99
: Attachment R0114, TVA Liquid Penetrant Examination, Reinspection Summary No. 01961-ISI-SQN,  for 2-SIH-020-IA, 5/19/14 R0105, TVA Liquid Penetrant Examination, Inspection Summary No. 01934-ISI-SQN2, or 2-CVCH-006-IA, 5/18/14
: NPG-SPP-09.1, ASME Code and Augmented Programs, Attachment 8, Form
: NPG-SPP-09.1-2, for Component ID, 2-SIH-020-IA, 5/19/14
: NPG-SPP-09.1, ASME Code and Augmented Programs, Attachment 8, Form
: NPG-SPP-09.1-2, for Component ID, 2-CVCH-585, 5/12/14 System 068, Reactor Coolant System Health Report, 2/1/2014 - 5/31/2014  
: 0-SI-DXI-000-114.3, Attachment 5, Unit-2 Examination Schedule for ASME Class 1, 2, 3
: Components, 5/9/14
: TVA Report No. R0105, Summary No. 01934-ISI-SQN2, Liquid Penetrant Examination Summary for Component ID 2-CVCH-006-IA, Category B-K/B10.20, Integral Attachment, 5/18/14
: Work Order No.
: 112354373, Valve
: SQN-2-VLV-001-0817 Replacement, 3/25/13 
: R0041, TVA Record of Visual Examination, 2-CVCH-585, 5/6/2014  
: R0086, TVA Record of Visual Examination, 2-CVCH-584, 5/15/2014 R0094, TVA Record of Visual Examination, 2-CVCH-586, 5/14/2014 R0151, Ultrasonic Piping Examination Data Sheet,
: FDF-011A, 5/24/2014  
: R0152, Ultrasonic Piping Examination Data Sheet,
: FDF-010C, 5/24/2014
: R0170, TVA Record of Visual Examination,
: SGH-4-1, 5/28/14
: Candidate No.
: 3237861, EPRI Performance Demonstration Initiative Program Qualifications,
: 1/14/11 MWK7861, IHI Southwest Technologies Inc. Certificate of Qualification, 2/22/2013
: H14132981, Certificate of Calibration, M&TE ID No. E41820, 4/14/2013
: VT-1, Certificate of Method Qualification Record for BMNO6QGPV, Expires, 11/28/2014
: VT-3, Certificate of Method Qualification Record for BMNO6QGPV, Expires, 11/28/2014
: VT-3, Certificate of Method Qualification Record for D880WSO0D, Expires, 10/5/2014 0-SI-MFT-000-001.0, Appendix "E" Page 1, Snubber Functional Testing,
: SQN-2-SNUB-015-SGBH104, 5/18/14 
: Report No.
: SCV-0001, Visual Examination of IWE Interfaces, Moisture Barrier, 5/20/2014  
: Report No.
: SCV-0004, Visual Examination of IWE Interfaces, Moisture Barrier, 5/5/2014
: Report No.
: SCV-0005, Visual Examination of IWE Interfaces, Moisture Barrier, 5/5/2014  
: 0-TI-DXX-000-097.1, Boric Acid Corrosion Control Program, Rev. 0009
: NPG-SPP-09.7.4, Boric Acid Corrosion Control Program, Rev. 0001 0-TI-SPT-000-301.0, ASME Section XI Pressure testing Program Basis Document, Rev. 0004
: 0-TI-RVI-000-301.0, Sequoyah Unit 1 & 2, PWR Reactor Vessel Internals Inspection Program, Rev. 0
: Sequoyah Unit 2 Control Room, Total Unidentified Leakage Logs, 12/28/2012 thru 5/10/2014
: Areva Use of Appendix H and Appendix I Qualified Techniques Sequoyah U2R19 Refueling Outage, Rev. 0, May 2014
: Areva 51-91988290-00, Sequoyah U2R19 Steam Generator Degradation Assessment,
: Rev. 0, May 2014
: Areva SQN 2C19 Analyst Training Instructions, Rev. 0
: Areva 54-ISI-400-021, Eddy Current Inspection Multi-Frequency Eddy Current Examination of Tubing, June 2013 Areva 51-9221442-000, Sequoyah Unit 2 EOC19 SG ECT Inspection Plan Areva ETSS_BOB1, Areva Examination Technical Specification Sheet for Bobbin Probe, Rev. 0 
: Attachment Areva ETSS_RPC1, Areva Examination Technical Specification Sheet for MRPC Probe, Rev. 0 Areva ETSS_RPC2, Areva Examination Technical Specification Sheet for MRPC Probe, Rev. 0 Areva ETSS_Array1, Areva Examination Technical Specification Sheet for MRPC Probe, Rev. 0 B85
: 130429 005, Sequoyah Unit 2 Refueling 18 Operational Assessment, Rev. 1
: Calibration Records for Eddy Current Tester Miz80i Serial Numbers:
: 39, 21, 71, 36, 73, 91
: Certificate of Conformance for Eddy Current Probes, Serial Numbers
: 653790,
: 652262,
: 655350,
: 653784,
: 652251,
: 653863,
: 652265,
: 652242 Calibration Standard for ASME 21095, 21099, 21100, 21096,
: EDM 9173936, 21086,
: ARRAY
: 9173939
: Personnel Qualification Records for Qualified Data Analysts:
: W. Bridforth, D. Cornell, 
: N. Farenbaugh, J. Janet' Sr, R. Lee, G. Manley, W. McMillan, S. Merriam, E. Miranda, 
: R. Miranda, J. Parrish, J. Oliver, A. Richardson, T. Shulter, J. Sordini, L. Tobin, D. Torres Personnel Qualification Records of TVA Steam Generator Program Personnel:
: J. Mayo,
: W. James
: SQN-ENG-F-10-02, Self-Assessment on Steam Generator Program, April 2010
: SQN-ENG-S-11-91, Benchmarking Report on U2R17 NRC Inservice Inspection Readiness, March 2011
: SQN-CEM-S-10-015, Self-Assessment on EPRI Secondary Water Chemistry Guidelines,
: July 2010
: Sequoyah Nuclear Plant Unit 2, Replacement Steam Generator Eddy Current Examination Guideline, Rev. 1
: Structural Integrity Associates, Report No. 1400660.401.R0, Independent Review of Westinghouse
: LTR-SGMMP-14-27, Assessment of Discolorations on Replacement Steam
: Generator Channel Head Cladding at Sequoyah Unit 2, dated May 30, 2014
: Section R12:
: Maintenance Effectiveness Procedures
: TI-4, Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting - 10CFR50.65, Revision 25
: Section R13:
: Maintenance Risk Assessments and Emergent Work Evaluation Procedures
: NPG-SPP-07.0, Work Management, Revision 0
: NPG-SPP-07.1, On Line Work Management, Revision 9
: NPG-SPP-07.3, Work Activity Risk Management Process, Revision 13
: NPG-SPP-07.11.1, Equipment Out of Service Management, Revision 9
: Section R15:
: Operability Evaluations Procedures
: NEDP-22, Functional Evaluations, Rev. 15
: OPDP-8, Limiting Conditions for Operation Tracking, Rev. 16
: NPG-SPP-03.5, Regulatory Reporting Requirements, Rev. 10
: Section R19:
: Post Maintenance Testing Procedures
: MMDP-1, Maintenance Management System, Rev. 20
: MMDP-3, Guidelines for Planning and Execution of Troubleshooting Activities, Rev. 6
: NPG-SPP-6.5, Foreign Material Control, Rev. 4 
: Attachment
: NPG-SPP-6.1, Work Order Process Initiation, Rev. 2
: NPG-SPP-06.3, Pre-/Post-Maintenance Testing, Rev. 1
: NPG-SPP-06.9, Testing Programs, Rev. 0
: NPG-SPP-06.9.1, Conduct of Testing, Rev. 8
: NPG-SPP-06.9.3, Post-Modification Testing, Rev. 5


==Section 2RS1: Radiological Hazard Assessment and Exposure Controls Procedures, Guidance Documents, and Manuals==
                                          7
: NPG-SPP-05.1 Radiological Controls Revision (Rev.) 003
Annual Inventory Reconciliation Confirmation for License #DRP-77, dated 01/14/2014
: NPG-SPP-05.1.1, Alpha Radiation Monitoring Program, Rev. 003
NPG Daily Outage Report, dated 05/12/2014
: NPG-SPP-05.6, Controlling Byproduct and Source Material, Rev. 002
NPG Daily Outage Report, dated 05/13/2014
: O-SI-RCI-000-056.0, Byproduct Material Inventory and Sealed Source Leak Test Rev. 016
RWP No. 14000063, LHRA - Plant Filter Change Outs: Seal Water Injection and Return, RCS,
: RCI-14, Radiation Work Permit (RWP) Program, Rev. 058
SFP, SFP Skimmer, Ion Exchange Filters and Refuel Water Purification Filters: Change Out and
: RCI-15, Radiological Postings Rev. 026
Transport.
: RCI-21, Control of Radioactive Materials, Rev. 019
RWP No. 14220052, U2 Lower Containment: IPCW - (HRAs) - MOVATs Testing
: RCI-22, Contamination Control Rev. 024
RWP No. 14220103, U2 Lower Containment, Excess LTDWN. HX. RM - LHRA
: RCI-24, Control of Very High Radiation Areas Rev. 014
RWP No. 14220122, U2 Lower Containment - Seal Table Work to Include Table Roll Back,
: RCI-28, Control of Locked High Radiation Areas Rev. 015
Tube Extraction, High Pressure Seals, Install Ferrules, Tube Cutting and Recovery Efforts
: RCI-29, Control of Radiation Protection Keys, Rev. 016
RWP No. 14230013, U2 Lower CTMT - Steam Generator Primary Platforms -LHRA
: RCI-201, Radiation and Contamination Surveys, Rev. 015
RWP No. 14230023, U2 Lower Containment - Steam Generator - LHRA - Full Jump for
: RCI-202, Airborne Radioactivity Surveys Rev. 008
Installing/Removing Nozzle Dams
: RCI-204, Radiological Surveys of Equipment and Materials Leaving the RCA, Rev. 008
Survey No. SQN-M-20140516-20, U2 Upper Containment - All Areas, 05/16/2014
: RCI-404, Radiation Protection Requirements for Remote Job Coverage, Rev. 001
Survey No. SQN-M-20140527-16, Reactor head move from cavity to head stand, 05/27/2014
: RCI-417, Radiological Monitoring of the Hydrogen Peroxide Injection Crud Burst, Rev. 001
Survey No. SQN-M-20140514-1, U-2 Raceway Elev 679, 05/14/2014
: RCDP-1, Conduct or Radiological Control Rev. 005
Survey No. SQN-M-20140515-4, U-2R19 Lower IPCW Floor Area, 05/15/2014
: Records and Data 0-SE-RCI-000-056.0 Byproduct Material Inventory and Sealed Source Leak Test, 11/23/2013 0-SE-RCI-000-056.0 Byproduct Material Inventory and Sealed Source Leak Test, 05/02/2014 0-TI-NUC-000-002.0, Storing Material in Spent Fuel Pool or New Fuel Vault, Rev. 0022, Attachment-1, Inventory of Materials in Spent Fuel Pool, Cask Loading Area, and New Fuel Vault, dated 02/06/2014.
Survey No. SQN-M-20140515-7, U2 #2 RCP Platform, 05/15/2014
: 2013 Sequoyah Radiation Protection Annual Report conducted per
Survey No. SQN-M-20140521-22, U2 Steam Generator (S/G) Platform, 05/21/2014
: NPG-SPP-05.1, Radiological Controls, Section 3.2, Program Monitoring Evaluation and Oversight Airborne Radiation Survey (ARS) Number (No.)
Survey No. SQN-M-20140522-6, U2R19 S/G 1&4 S/G Generator Platform, 05/22/2014
: 051514008, U-2 Lower Seal Table During Eddy Current Test, dated 05/15/2014
Survey No. SQN-M-20140605-4, 5&6, SQN ISFSI PAD Quarterly Routine, 06/05/2014
: ARS No.
Survey No. SQN-M-2014021-3, 5&6, SQN ISFSI PAD Quarterly Routine, 02/16/2014
: 051514004, U-2 Lower Containment Raceway Routine, dated 05/15/2014 
Survey No. SQN-M-20131020-15, SQN ISFSI PAD Quarterly Routine, 10/20/2013
: ARS No.
Corrective Action Program (CAP) Documents
: 051514002, U-2 Lower Containment Routine @Elevation 679, dated 05/15/2014
PER-661017
: ARS No.
PER-713213
: 051514003, U-2 Lower Containment Routine @IPCW, dated 05/15/2014 ARS No.
PER-776043
: 052114017, U-2 S/G#1 Laydown Area During Insert Removal, dated 05/21/2014 ARS No.
PER-776044
: 052114018, U-2 S/G#1 During Insert Removal, dated 05/21/2014
PER-790597
: ARS No.
PER 805944
: 052114019, U-2 S/G#4 During Insert Removal, dated 05/21/2014
PER-805952
: ARS No.
PER 807919
: 052114022, U-2 S/G#3 During Insert Removal, dated 05/21/2014
PER 827948
: ARS No.
PER 868727
: 052114023, U-2 S/G#2 During Insert Removal, dated 05/21/2014
PER 881321
: ARS No.
PER 886160
: 052114024, U-2 S/G#3 Laydown Area During Insert Removal, dated 05/21/2014 ARS No.
PER 886668
: 052114025, U-2 S/G#1 Laydown Area Back-up Sample, dated 05/21/2014 ARS No.
PER 888770
: 052114026, U-2 S/G#3 Laydown Area Back-up Sample, dated 05/21/2014
Section 2RS2: Occupational ALARA Planning and Controls
: ARS No.
Procedures, Guidance Documents, and Manuals
: 052114027, U-2 S/G#3 Primary PlatformBack-up Sample, dated 05/21/2014 
CHEM -002, Primary Water Chemistry Program Strategic Plan, Rev. 6
: Attachment Annual Inventory Reconciliation Confirmation for License #DRP-77, dated 01/14/2014 NPG Daily Outage Report, dated 05/12/2014 NPG Daily Outage Report, dated 05/13/2014  
NPG-SPP-05.2, ALARA Program, Rev. 4
: RWP No.
NPG-SPP-05.2.1, Operational ALARA Planning and Controls, Rev. 2
: 14000063, LHRA - Plant Filter Change Outs: Seal Water Injection and Return, RCS,  
NPG-SPP-05.2.2, Establishing Collective Radiation Exposure Annual Business Plan Goals,
: SFP, SFP Skimmer, Ion Exchange Filters and Refuel Water Purification Filters: Change Out and Transport.
Rev. 0
: RWP No.
NPG-SPP-05.2.3, Outage Exposure Estimating and Tracking, Rev. 0
: 14220052, U2 Lower Containment: IPCW - (HRA's) - MOVATs Testing RWP No.
                                                                                  Attachment
: 14220103, U2 Lower Containment, Excess LTDWN. HX. RM - LHRA
: RWP No.
: 14220122, U2 Lower Containment - Seal Table Work to Include Table Roll Back, Tube Extraction, High Pressure Seals, Install Ferrules, Tube Cutting and Recovery Efforts  
: RWP No.
: 14230013, U2 Lower CTMT - Steam Generator Primary Platforms -LHRA  
: RWP No.
: 14230023, U2 Lower Containment - Steam Generator - LHRA - Full Jump for Installing/Removing Nozzle Dams Survey No.
: SQN-M-20140516-20, U2 Upper Containment - All Areas, 05/16/2014  
: Survey No.
: SQN-M-20140527-16, Reactor head move from cavity to head stand, 05/27/2014  
: Survey No.
: SQN-M-20140514-1, U-2 Raceway Elev 679, 05/14/2014  
: Survey No.
: SQN-M-20140515-4, U-2R19 Lower IPCW Floor Area, 05/15/2014 Survey No.
: SQN-M-20140515-7, U2 #2 RCP Platform, 05/15/2014 Survey No.
: SQN-M-20140521-22, U2 Steam Generator (S/G) Platform, 05/21/2014  
: Survey No.
: SQN-M-20140522-6, U2R19 S/G 1&4 S/G Generator Platform, 05/22/2014  
: Survey No.
: SQN-M-20140605-4, 5&6, SQN ISFSI PAD Quarterly Routine, 06/05/2014  
: Survey No.
: SQN-M-2014021-3, 5&6, SQN ISFSI PAD Quarterly Routine, 02/16/2014  
: Survey No.
: SQN-M-20131020-15, SQN ISFSI PAD Quarterly Routine, 10/20/2013
: Corrective Action Program (CAP) Documents
: PER-661017
: PER-713213
: PER-776043
: PER-776044
: PER-790597
: PER 805944
: PER-805952
: PER 807919
: PER 827948
: PER 868727
: PER 881321
: PER 886160
: PER 886668  
: PER 888770


==Section 2RS2: Occupational==
                                          8
: ALARA Planning and Controls Procedures, Guidance Documents, and Manuals CHEM -002, Primary Water Chemistry Program Strategic Plan, Rev. 6
RCI-10, ALARA Program, Rev. 35
: NPG-SPP-05.2, ALARA Program, Rev. 4
RCI-14, Radiation Work Permit (RWP) Program, Rev. 58
: NPG-SPP-05.2.1, Operational ALARA Planning and Controls, Rev. 2
RCI-19, Temporary Shielding Program, Rev. 13
: NPG-SPP-05.2.2, Establishing Collective Radiation Exposure Annual Business Plan Goals, Rev. 0
RCI-417, Radiological Monitoring of the Hydrogen Peroxide Injection Crud Burst, Rev. 1
: NPG-SPP-05.2.3, Outage Exposure Estimating and Tracking, Rev. 0 
Reports, Records, and Data
: Attachment
ALARA Committee Meeting Minutes - Meeting Number (No.) 2013-04, 2/22/2013
: RCI-10, ALARA Program, Rev. 35
ALARA Committee Meeting Minutes - Meeting No. 2013-11, 7/11/2013
: RCI-14, Radiation Work Permit (RWP) Program, Rev. 58
ALARA Committee Meeting Minutes - Meeting No. 2013-19, 10/17/ 2013
: RCI-19, Temporary Shielding Program, Rev. 13
ALARA Committee Meeting Minutes - Meeting No. 2013-22, 11/14/ 2013
: RCI-417, Radiological Monitoring of the Hydrogen Peroxide Injection Crud Burst, Rev. 1  
ALARA Plan: 2013-010, Refueling Operations
: Reports, Records, and Data ALARA Committee Meeting Minutes - Meeting Number (No.) 2013-04, 2/22/2013 ALARA Committee Meeting Minutes - Meeting No. 2013-11, 7/11/2013 ALARA Committee Meeting Minutes - Meeting No. 2013-19, 10/17/ 2013  
ALARA Plan: 2013-011, Mechanical Maintenance Group (MMG)
: ALARA Committee Meeting Minutes - Meeting No. 2013-22, 11/14/ 2013  
ALARA Plan: 2013-012, Electrical Maintenance and RCPs
: ALARA Plan: 2013-010, Refueling Operations  
ALARA Plan: 2013-015, Plant Services
: ALARA Plan: 2013-011, Mechanical Maintenance Group (MMG) ALARA Plan: 2013-012, Electrical Maintenance and RCPs ALARA Plan: 2013-015, Plant Services  
ALARA Plan: 2013-017, Radiation Protection
: ALARA Plan: 2013-017, Radiation Protection  
ALARA Plan: 2013-018, U1R19 MODS Ice Condenser/Snubbers/Insulation/Scaffolds/Painting
: ALARA Plan: 2013-018, U1R19 MODS Ice Condenser/Snubbers/Insulation/Scaffolds/Painting  
ALARA Plan: 2014-010, Refueling Operations
: ALARA Plan: 2014-010, Refueling Operations ALARA Plan: 2014-011, Mechanical Maintenance Group (MMG) ALARA Plan: 2014-015, Plant Services (RCL)  
ALARA Plan: 2014-011, Mechanical Maintenance Group (MMG)
: ALARA Plan: 2014-017, Radiation Protection  
ALARA Plan: 2014-015, Plant Services (RCL)
: ALARA Plan: 2013-018, Modifications U2R19  
ALARA Plan: 2014-017, Radiation Protection
: ALARA Work in Progress Review:
ALARA Plan: 2013-018, Modifications U2R19
: RWP 2013-011, 10/24/13  
ALARA Work in Progress Review: RWP 2013-011, 10/24/13
: ALARA Work in Progress Review:
ALARA Work in Progress Review: RWP 2013-018, 11/14/13
: RWP 2013-018, 11/14/13 Fiscal Year (FY) Dose Estimate Approval per
Fiscal Year (FY) Dose Estimate Approval per NPG-SPP-5.2.2 3.1.2
: NPG-SPP-5.2.2 3.1.2 FY14 RP Dose Reduction Plan Graphic Showing Co-58 and Co-60 Trends for 2013-2024 in Both Units Report Sequoyah TEDE Year to Date as of 4/24/14  
FY14 RP Dose Reduction Plan
: SQN U1R19 Outage - Dose Reduction Plan Slide Show: Sequoyah Nuclear Plant 2011-2015 Business Plan:
Graphic Showing Co-58 and Co-60 Trends for 2013-2024 in Both Units
: Collective Radiation Exposure
Report Sequoyah TEDE Year to Date as of 4/24/14
: 2014-2018 SQN Business Plan  
SQN U1R19 Outage - Dose Reduction Plan
: U1R19 ALARA OUTAGE REPORT  
Slide Show: Sequoyah Nuclear Plant 2011-2015 Business Plan: Collective Radiation Exposure
: U1R19 Final ALARA Plan Status  
2014-2018 SQN Business Plan
: U2R19 ALARA Plan Challenge Numbers Spreadsheet  
U1R19 ALARA OUTAGE REPORT
: CAP Documents 2013 Sequoyah Radiation Protection Annual Report, 10/25/2013  
U1R19 Final ALARA Plan Status
: Audit SSA1309, Radiation Protection Sequoyah Nuclear Plant, 8/19-30/2013
U2R19 ALARA Plan Challenge Numbers Spreadsheet
: PER 773873
CAP Documents
: PER 776064
2013 Sequoyah Radiation Protection Annual Report, 10/25/2013
: PER 776639
Audit SSA1309, Radiation Protection Sequoyah Nuclear Plant, 8/19-30/2013
: PER 770709
PER 773873
: PER 773258
PER 776064
: PER 724010
PER 776639
: PER 798963
PER 770709
: PER 801067
PER 773258
: PER 886820
PER 724010
: PER 853897  
PER 798963
: Attachment
PER 801067
PER 886820
PER 853897
                                                                                  Attachment


==Section 2RS3: In-Plant Airborne==
                                          9
: Radioactivity Control And Mitigation Procedures and Guidance Documents 0-PI-FPU-049-401.M, Self Contained Breathing Apparatus, Rev. 030  
Section 2RS3: In-Plant Airborne Radioactivity Control And Mitigation
: 0-PI-RCI-033-001.0, Periodic Monitoring of Service Air System for Use as Breathing Air, Rev. 008
Procedures and Guidance Documents
: NPG-SPP-05.10, Radiological Respiratory Protection Program, Rev. 003
0-PI-FPU-049-401.M, Self Contained Breathing Apparatus, Rev. 030
: RCI-04.01, Selection, Issue, and Use of Respiratory Protection Devices, Rev. 008
0-PI-RCI-033-001.0, Periodic Monitoring of Service Air System for Use as Breathing Air,
: RCI-04.02, Cleaning/Sanitizing, Maintenance, Inspection, Storage and Inventory or Respiratory Protection Devices, Rev 4
Rev. 008
: RCI-04.03, Respiratory Protection Program Periodic Evaluation Rev. 000
NPG-SPP-05.10, Radiological Respiratory Protection Program, Rev. 003
: RCI-18.01, DOP Testing of Portable HEPA and Vacuum Cleaners, Rev 001  
RCI-04.01, Selection, Issue, and Use of Respiratory Protection Devices, Rev. 008
: UFSAR Chapter 11 & 12
RCI-04.02, Cleaning/Sanitizing, Maintenance, Inspection, Storage and Inventory or Respiratory
: Records and Data Reviewed
Protection Devices, Rev 4
: AIR/GAS Quality Report and Certificates for SN: 11040, Kit #279317, dated 10/01/2013 and Kit  
RCI-04.03, Respiratory Protection Program Periodic Evaluation Rev. 000
#
RCI-18.01, DOP Testing of Portable HEPA and Vacuum Cleaners, Rev 001
: 286878, dated 03/20/2013  
UFSAR Chapter 11 & 12
: Assessment
Records and Data Reviewed
: SQN-RP-S-14-003, In-Plant Airborne Radioactivity Control and Mitigation, 01/15/2014 Grade D Certificates for Plant System Air Compressor Equipment ID#s 0-CLR-32-25, 0-CLR-
AIR/GAS Quality Report and Certificates for SN: 11040, Kit #279317, dated 10/01/2013 and Kit
: 2-26, 0-CLR-32-27 and 0-DS-32-136, dated 09/25/2013  
# 286878, dated 03/20/2013
: Grade D Certificates for Plant System Air Compressor Equipment ID#s 0-CLR-32-25, 0-CLR-
Assessment SQN-RP-S-14-003, In-Plant Airborne Radioactivity Control and Mitigation,
: 2-26, 0-CLR-32-27 and 0-DS-32-136, dated 02/13/2012  
01/15/2014
: HEPA DOP Test Certification for Vacuum Cleaner #s
Grade D Certificates for Plant System Air Compressor Equipment ID#s 0-CLR-32-25, 0-CLR-
: TVA-2 and 1369, dated 05/14/2014
32-26, 0-CLR-32-27 and 0-DS-32-136, dated 09/25/2013
: KeyStone Certifications for Vacuum HEPA Filters #2801, #2787 and #2790, dated 03/13/2014
Grade D Certificates for Plant System Air Compressor Equipment ID#s 0-CLR-32-25, 0-CLR-
: MSA MMR Certification Records for TVA SCBA Repair Technicians, Current
32-26, 0-CLR-32-27 and 0-DS-32-136, dated 02/13/2012
: MSA Posi3 USB Complete SCBA Test Results for Units CR06, CR12 and CR13, dated  
HEPA DOP Test Certification for Vacuum Cleaner #s TVA-2 and 1369, dated 05/14/2014
: 08/06/2013  
KeyStone Certifications for Vacuum HEPA Filters #2801, #2787 and #2790, dated 03/13/2014
: Personnel Contamination Log, 1/2013-5/2014 UNITECH Services Group DOP Test Results for HEPA #700-7, dated 12/09/2014 UNITECH Services Group DOP Test Results for HEPA #700-8, dated 11/01/2013  
MSA MMR Certification Records for TVA SCBA Repair Technicians, Current
: UNITECH Services Group DOP Test Results for HEPA #700-29, dated 10/02/2013  
MSA Posi3 USB Complete SCBA Test Results for Units CR06, CR12 and CR13, dated
: CAP Documents
08/06/2013
: PER 660950 PER 805989
Personnel Contamination Log, 1/2013-5/2014
UNITECH Services Group DOP Test Results for HEPA #700-7, dated 12/09/2014
UNITECH Services Group DOP Test Results for HEPA #700-8, dated 11/01/2013
UNITECH Services Group DOP Test Results for HEPA #700-29, dated 10/02/2013
CAP Documents
PER 660950
PER 805989
Section 2RS4: Occupational Dose Assessment
Procedures and Guidance Documents
NPG-SPP-05.1.1 Alpha Radiation Monitoring Program, Rev. 003
RCI-05.304, WBC Routine Operations and DAC-Hr Assignment Evaluation, Rev. 009
RCI-202, Airborne Radioactivity Surveys, Rev. 008
RCI-209, Radiological Surveys of Personnel Leaving the RCA or Protected Area, Rev. 004
RCDP-7, Bioassay and Internal Dose Program, Rev. 005
RCDP-10, Personnel Contamination Reporting, Rev. 005
RCTP-106 Special Dosimetry Operations, Rev. 003
RCTP-113, External Dosimetry MQA Program, Rev. 000
                                                                                  Attachment


==Section 2RS4: Occupational Dose Assessment Procedures and Guidance Documents==
                                          10
: NPG-SPP-05.1.1 Alpha Radiation Monitoring Program, Rev. 003
Records and Data Reviewed
: RCI-05.304, WBC Routine Operations and
Assessment SQN-RP-S-14-004, Occupational Dose Assessment, 02/10/2014
: DAC-Hr Assignment Evaluation, Rev. 009
Committed Effective Dose Equivalent Assignment Summary for 2013
: RCI-202, Airborne Radioactivity Surveys, Rev. 008
Dosimetry Investigation Reports 2014-015, 2014-016 and 2014-017
: RCI-209, Radiological Surveys of Personnel Leaving the RCA or Protected Area, Rev. 004
Evaluation of the Canberra GEM-5 Portal Contamination Monitor Detection Capabilities for Use
: RCDP-7, Bioassay and Internal Dose Program, Rev. 005
as a Passive Whole Body Count Instrument, dated 10/30/2014
: RCDP-10, Personnel Contamination Reporting, Rev. 005
Investigative Whole body Counts (6) for Intakes Occurring on U2 S/G Platform on 05/21/2014
: RCTP-106 Special Dosimetry Operations, Rev. 003
Multi-Badge EDEX Worksheet for Entry on RWP No. 14240182, dated 05/15/2014
: RCTP-113, External Dosimetry MQA Program, Rev. 000 
Multi-Badge EDEX Worksheet for Entry on RWP No. 14240213, dated 05/14/2014
: Attachment Records and Data Reviewed
Multi-Badge EDEX Worksheet for Entry on RWP No. 14240053, dated 05/14/2014
: Assessment
Multi-Badge EDEX Worksheets (2) for Entries on RWP No. 14240023, dated 05/23/2014 and
: SQN-RP-S-14-004, Occupational Dose Assessment, 02/10/2014 Committed Effective Dose Equivalent Assignment Summary for 2013  
05/29/2014
: Dosimetry Investigation Reports 2014-015, 2014-016 and 2014-017  
NVLAP Certification of Accreditation to ISO/IEC 17025-2005 for 2014
: Evaluation of the Canberra
OSL Dosimetry Investigation Summary for 01/2013-05/2014
: GEM-5 Portal Contamination Monitor Detection Capabilities for Use as a Passive Whole Body Count Instrument, dated 10/30/2014  
SQN TLD Area Monitoring Results for 4th Qtr. 2013
: Investigative Whole body Counts (6) for Intakes Occurring on U2 S/G Platform on 05/21/2014
CAP Documents
: Multi-Badge EDEX Worksheet for Entry on RWP No.
PER 675250
: 14240182, dated 05/15/2014
PER 753263
: Multi-Badge EDEX Worksheet for Entry on RWP No.
PER 784430
: 14240213, dated 05/14/2014
PER 798104
: Multi-Badge EDEX Worksheet for Entry on RWP No.
PER 829995
: 14240053, dated 05/14/2014  
PER 830008
: Multi-Badge EDEX Worksheets (2) for Entries on RWP No.
PER 845120
: 14240023, dated 05/23/2014 and  
PER 857054
: 05/29/2014 NVLAP Certification of Accreditation to ISO/IEC 17025-2005 for 2014 OSL Dosimetry Investigation Summary for 01/2013-05/2014  
PER 869683
: SQN TLD Area Monitoring Results for 4th Qtr. 2013
PER 881323
: CAP Documents
PER 888629
: PER 675250
PER 888987
: PER 753263
Section 2RS5: Radiation Monitoring Instrumentation
: PER 784430
Procedures and Guidance Documents
: PER 798104
1-SI-ICC-090-400.0, Calibration of Shield Building Vent Radiation Monitor 1-RM-90-400,
: PER 829995
Rev. 18
: PER 830008
2-SI-ICC-090-400.0, Calibration of Shield Building Vent Radiation Monitor 2-R-90-400, Rev. 18
: PER 845120
CHTP-109, Chemistry QA/QC, Rev. 8
: PER 857054
EPIP-1, Emergency Plan Classification Matrix, Rev. 50
: PER 869683
NPG-SPP-06.7, Instrumentation Setpoint, Scaling and Calibration Program, Rev. 2
: PER 881323
RCI-5, Radiation Protection Instrumentation Program, Rev.77
: PER 888629 PER 888987
RCI-5.100, Operation of Laboratory Counter/ Scalers, Rev. 6
RCI-5.102, Calibration and Operation of the Canberra iSOLO Model 300G Alpha/Beta Counter,
Rev. 5
RCI-5.300, Calibration and Operation of the Eberline Personnel Contamination Monitor (PCM-
1B), Rev. 3
RCI-05-301, Operational Checks for the GEM-5 Portal Monitor, Rev. 8
RCI-05.305, Calibration, Response Check, And Operation of the Canberra ARGOS-5AB
Personnel Contamination Monitor, Rev.7
                                                                                  Attachment


==Section 2RS5: Radiation Monitoring Instrumentation Procedures and Guidance Documents==
                                            11
: 1-SI-ICC-090-400.0, Calibration of Shield Building Vent Radiation Monitor 1-RM-90-400, Rev. 18 2-SI-ICC-090-400.0, Calibration of Shield Building Vent Radiation Monitor 2-R-90-400, Rev. 18
RCI-05.306, Calibration, Response Check, and Operation of the Canberra Cronos-4and
: CHTP-109, Chemistry QA/QC, Rev. 8
Cronos-11 Contamination Monitors, Rev.3
: EPIP-1, Emergency Plan Classification Matrix, Rev. 50
RCI-05.400, Criteria for Setting Portable Radiation Protection Instrument Response Check
: NPG-SPP-06.7, Instrumentation Setpoint, Scaling and Calibration Program, Rev. 2
Windows, Rev.4
: RCI-5, Radiation Protection Instrumentation Program, Rev.77
RCI-05.408, Response Check of Neutron Survey Instruments, Rev. 0
: RCI-5.100, Operation of Laboratory Counter/ Scalers, Rev. 6
RCI-05.401, Instrument Response Checks Utilizing the Shepherd Calibrator, Rev.4
: RCI-5.102, Calibration and Operation of the Canberra iSOLO Model 300G Alpha/Beta Counter, Rev. 5
Records and Data Reviewed
: RCI-5.300, Calibration and Operation of the Eberline Personnel Contamination Monitor (PCM-
Apex Gamma Spectroscopy Efficiency Calibration, Detector 2, 1/24/2012
: 1B), Rev. 3
Calibration Data Records for the following instruments:
: RCI-05-301, Operational Checks for the
ARGOS-5AB, TVA# 860588, 5/8/2013 and 4/30/2014
: GEM-5 Portal Monitor, Rev. 8
ARGOS-5AB, TVA# 860589, 4/26/2013 and 4/9/2014
: RCI-05.305, Calibration, Response Check, And Operation of the Canberra
Bicron Analyst [no probe type specified], TVA# 8355305, 7/30/2013
: ARGOS-5AB Personnel Contamination Monitor, Rev.7
Bicron Analyst with GM, TVA# 8355305, 2/4/2014
: Attachment
Bicron Analyst with NaI, TVA# 835539, 6/4/2103 and 1/7/2014
: RCI-05.306, Calibration, Response Check, and Operation of the Canberra Cronos-4and Cronos-11 Contamination Monitors, Rev.3
Cronos, TVA# 860780, S/N 1203-021, 2/5/2013 and 1/27/2014
: RCI-05.400, Criteria for Setting Portable Radiation Protection Instrument Response Check Windows, Rev.4
Eberline Teletector, TVA# 523331, 8/28/2013 and 2/26/2014
: RCI-05.408, Response Check of Neutron Survey Instruments, Rev. 0
Eberline Teletector, TVA# 523338, 8/22/2013 and 2/26/2014
: RCI-05.401, Instrument Response Checks Utilizing the Shepherd Calibrator, Rev.4  
GEM-5, S/N 1203-021, 5/7/2013 and 4/23/2014
: Records and Data Reviewed
GEM-5, S/N 0909-179, 3/18/2013 and 3/14/2014
: Apex Gamma Spectroscopy Efficiency Calibration, Detector 2, 1/24/2012  
HV-1 [air sampler], TVA# 556318, 10/1/2013 and 3/24/2014
: Calibration Data Records for the following instruments:
HV-1 [air sampler], TVA# 860003, 8/30/2013 and 3/27/2014
: ARGOS-5AB, TVA#
Ludlum Model 3 frisker, TVA# 860888, 5/30/2013 and 2/18/2014
: 860588, 5/8/2013 and 4/30/2014
Ludlum 2200, TVA# 860654, 10/17/2012 and 12/3/2013
: ARGOS-5AB, TVA#
Ludlum 3030P, TVA# 951047, 12/17/2014
: 860589, 4/26/2013 and 4/9/2014 Bicron Analyst [no probe type specified], TVA#
Ludlum 9-3, TVA# 860844, 5/22/2013 and 2/4/2014
: 8355305, 7/30/2013 Bicron Analyst with GM, TVA#
Ludlum 9-3, TVA# 861000, 2/25/2013 and 2/5/2014
: 8355305, 2/4/2014  
MG Telepole WR, TVA# 860096, 3/5/2013 and 1/7/2014
: Bicron Analyst with NaI, TVA#
MG Telepole WR, TVA# 951056, 3/3/2014 and 5/14/2014
: 835539, 6/4/2103 and 1/7/2014  
SAIC H-810, TVA# 838786 and TVA# 860029, 6/5/2013 and 2/24/2014
: Cronos, TVA#
SAM-11, TVA# 860323, 6/17/2013 and 3/10/2014
: 860780, S/N 1203-021, 2/5/2013 and 1/27/2014  
SAM-11, TVA# 860324, 2/4/2013 and 1/30/2014
: Eberline Teletector, TVA#
PCM-1B, TVA# 576358, 8/16/2012 and 7/26/2013
: 523331, 8/28/2013 and 2/26/2014 Eberline Teletector, TVA#
PCM-1B, TVA# 484689, 4/19/2012 and 9/23/2013
: 523338, 8/22/2013 and 2/26/2014
Calibration/Efficiency Check, Tri-Carb Model 3100TR, S/N 060450, 6/2/2014
: GEM-5, S/N 1203-021, 5/7/2013 and 4/23/2014
Calibration Report, Calibration of the FASTSCAN 1 WBC System at the Dosimetry Lab of the
: GEM-5, S/N 0909-179, 3/18/2013 and 3/14/2014
TVA Sequoyah Nuclear Plant, 8/22/2013
: HV-1 [air sampler], TVA#
Calibration Report, Calibration of the FASTSCAN 2 WBC System at the Dosimetry Lab of the
: 556318, 10/1/2013 and 3/24/2014
TVA Sequoyah Nuclear Plant, 8/22/2013
: HV-1 [air sampler], TVA#
Calibration Report, High Range Radiation Monitor Calibrator RT-11, S/N 24, TVA Source #
: 860003, 8/30/2013 and 3/27/2014  
775N, 3/15/1983
: Ludlum Model 3 frisker, TVA#
Certificate of Calibration, Beta Standard Source, S/N G4-973, TVA Source # 2482, 11/15/2009
: 860888, 5/30/2013 and 2/18/2014 Ludlum 2200, TVA#
Certificate of Calibration, Beta Standard Source, S/N G4-972, TVA Source # 2485, 11/15/2009
: 860654, 10/17/2012 and 12/3/2013 Ludlum 3030P, TVA#
Certificate of Calibration, Standard Radionuclide Source, S/N 86078-166, 10/1/2011
: 951047, 12/17/2014  
Certificate of Calibration, Standard Radionuclide Source, S/N 95460, 1/1/2014
: Ludlum 9-3, TVA#
Certificate of Gamma Standard Source, S/N 205-56-4, TVA Source # 1295N, 8/1/1989
: 860844, 5/22/2013 and 2/4/2014  
Certificate of Gamma Standard Source, S/N 363-02-3, TVA Source # 1296N, 5/1/1990
: Ludlum 9-3, TVA#
Certificate of Gamma Standard Source, S/N 205-83-5, TVA Source # 1297N, 1/1/1990
: 861000, 2/25/2013 and 2/5/2014  
                                                                                    Attachment
: MG Telepole WR, TVA#
: 860096, 3/5/2013 and 1/7/2014 MG Telepole WR, TVA#
: 951056, 3/3/2014 and 5/14/2014 SAIC H-810, TVA#
: 838786 and TVA#
: 860029, 6/5/2013 and 2/24/2014
: SAM-11, TVA#
: 860323, 6/17/2013 and 3/10/2014
: SAM-11, TVA#
: 860324, 2/4/2013 and 1/30/2014
: PCM-1B, TVA#
: 576358, 8/16/2012 and 7/26/2013
: PCM-1B, TVA#
: 484689, 4/19/2012 and 9/23/2013
: Calibration/Efficiency Check, Tri-Carb Model 3100TR, S/N
: 060450, 6/2/2014 Calibration Report, Calibration of the "FASTSCAN 1" WBC System at the Dosimetry Lab of the  
: TVA Sequoyah Nuclear Plant, 8/22/2013  
: Calibration Report, Calibration of the "FASTSCAN 2" WBC System at the Dosimetry Lab of the  
: TVA Sequoyah Nuclear Plant, 8/22/2013 Calibration Report, High Range Radiation Monitor Calibrator
: RT-11, S/N 24, TVA Source # 775N, 3/15/1983  
: Certificate of Calibration, Beta Standard Source, S/N G4-973, TVA Source # 2482, 11/15/2009  
: Certificate of Calibration, Beta Standard Source, S/N G4-972, TVA Source # 2485, 11/15/2009  
: Certificate of Calibration, Standard Radionuclide Source, S/N 86078-166, 10/1/2011  
: Certificate of Calibration, Standard Radionuclide Source, S/N 95460, 1/1/2014 Certificate of Gamma Standard Source, S/N 205-56-4, TVA Source # 1295N, 8/1/1989 Certificate of Gamma Standard Source, S/N 363-02-3, TVA Source # 1296N, 5/1/1990  
: Certificate of Gamma Standard Source, S/N 205-83-5, TVA Source # 1297N, 1/1/1990
: Attachment Certificate of Gamma Standard Source, S/N M-246, TVA Source # 1297N, 5/15/1990 Certificate of Gamma Standard Source, S/N M-250, TVA Source # 1299, 5/15/1990 Certificate of Gamma Standard Source, S/N M-248, TVA Source # 1300N, 5/15/1990
: Certificate of Gamma Standard Source, S/N 349-29-1, TVA Source # 1301, 8/1/1989
: Digital Air Flow Calibrator, TVA#
: 860169, S/N 3204, Source Check Record, 11/23/13
: F & J Specialty Products, Inc. Certificate of Calibration, Digital Calibrator Model D-828B, Serial
# (S/N) 3204, 11/5/2013 Sequoyah Offsite Dose Calculation Manual (ODCM), Rev. 58 Source Response and Background Data Sheet, Ludlum 2200 Scaler, TVA#
: 860654, February and March 2014
: System Health Report, System 90, Radiation Monitoring, 10/1/2013 through 1/31/2014 White Paper, Waste Stream Analysis (DAW 10/14/2013), 3/15/2014
: White Paper, Sequoyah Whole Body Counter Library Revision, 5/30/2014 Whole Body Counter Library Listing, Europium -152, 6/4/2014
: Work Order (WO) No.
: 112807041, 1-SI-ICC-090-400.0 Shield Building Vent Rad Mon 1-RM-90-
: 400 Cal, 10/10/2012
: WO No.
: 115052959, 1-PI-CEM-043-487.0 U1 Post Accident Sampling Sys Calibration, 1/29/2014 WO No.
: 112625727, 0-SI-ICC-090-101.B Aux Bldg Vent Gaseous Rad Mon 0-R-90-101B &
: Flow Monitor 0-F-30-174 CC, 8/31/2012
: WO No.
: 114475841, 0-SI-ICC-090-101.B Aux Bldg Vent Gaseous Rad Mon 0-R-90-101B &
: Flow Monitor 0-F-30-174 CC, 9/6/2013
: CAP Documents Assessment
: SQN-RP-S-14-002, RP Portable Instrumentation and Calibration, 1/28 to 1/30/2014
: PER 735601
: PER793878
: PER 801879
: PER 832856 PER 871912


==Section 4OA1: Performance Indicator Verification Procedures==
                                            12
: NPG-SPP-02.2, Performance Indicator Program, Rev. 2
Certificate of Gamma Standard Source, S/N M-246, TVA Source # 1297N, 5/15/1990
: NPG-SPP-02.2, Performance Indicator Program, Rev. 6
Certificate of Gamma Standard Source, S/N M-250, TVA Source # 1299, 5/15/1990
: NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Rev. 6  
Certificate of Gamma Standard Source, S/N M-248, TVA Source # 1300N, 5/15/1990
: Records and Data Reviewed 2013 Annual Radioactive Effluent Release Report, 4/16/2014 2013 Annual Radiological Environmental Operating Report, 4/16/2014 Gaseous Radioactive Waste Release Permit 2014027.059.014.G  
Certificate of Gamma Standard Source, S/N 349-29-1, TVA Source # 1301, 8/1/1989
: Liquid Radioactive Waste Release Permit 2013191.007.087.L  
Digital Air Flow Calibrator, TVA# 860169, S/N 3204, Source Check Record, 11/23/13
: CAP Documents
F & J Specialty Products, Inc. Certificate of Calibration, Digital Calibrator Model D-828B, Serial
: PER 756809
# (S/N) 3204, 11/5/2013
: PER 824084
Sequoyah Offsite Dose Calculation Manual (ODCM), Rev. 58
: PER 788604
Source Response and Background Data Sheet, Ludlum 2200 Scaler, TVA# 860654, February
: PER 793921
and March 2014
: Attachment
System Health Report, System 90, Radiation Monitoring, 10/1/2013 through 1/31/2014
White Paper, Waste Stream Analysis (DAW 10/14/2013), 3/15/2014
White Paper, Sequoyah Whole Body Counter Library Revision, 5/30/2014
Whole Body Counter Library Listing, Europium -152, 6/4/2014
Work Order (WO) No. 112807041, 1-SI-ICC-090-400.0 Shield Building Vent Rad Mon 1-RM-90-
400 Cal, 10/10/2012
WO No. 115052959, 1-PI-CEM-043-487.0 U1 Post Accident Sampling Sys Calibration,
1/29/2014
WO No. 112625727, 0-SI-ICC-090-101.B Aux Bldg Vent Gaseous Rad Mon 0-R-90-101B &
Flow Monitor 0-F-30-174 CC, 8/31/2012
WO No. 114475841, 0-SI-ICC-090-101.B Aux Bldg Vent Gaseous Rad Mon 0-R-90-101B &
Flow Monitor 0-F-30-174 CC, 9/6/2013
CAP Documents
Assessment SQN-RP-S-14-002, RP Portable Instrumentation and Calibration, 1/28 to 1/30/2014
PER 735601
PER793878
PER 801879
PER 832856
PER 871912
Section 4OA1: Performance Indicator Verification
Procedures
NPG-SPP-02.2, Performance Indicator Program, Rev. 2
NPG-SPP-02.2, Performance Indicator Program, Rev. 6
NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Rev. 6
Records and Data Reviewed
2013 Annual Radioactive Effluent Release Report, 4/16/2014
2013 Annual Radiological Environmental Operating Report, 4/16/2014
Gaseous Radioactive Waste Release Permit 2014027.059.014.G
Liquid Radioactive Waste Release Permit 2013191.007.087.L
CAP Documents
PER 756809
PER 824084
PER 788604
PER 793921
                                                                                        Attachment


==Section 4OA5: Temporary Instruction 2515/182 - Review of the Industry Initiative to Control Degradation of Underground Piping and Tanks Corrective Action Program Documents==
                                            13
: PER 175149-002, 4" Diesel Fuel Oil Line Failed Pressure Test
Section 4OA5: Temporary Instruction 2515/182 - Review of the Industry Initiative to
: PER 347970,
Control Degradation of Underground Piping and Tanks
: NEI 09-14, NSIAC Buried Piping Completion Dates To Be Assigned as LTCAs
Corrective Action Program Documents
: PER 684460, License Renewal NRC Commitment #3: Revise the Buried and Underground Piping and Tanks Inspection Program to Meet License Renewal Requirements  
PER 175149-002, 4 Diesel Fuel Oil Line Failed Pressure Test
===Procedures===
PER 347970, NEI 09-14, NSIAC Buried Piping Completion Dates To Be Assigned as LTCAs
: 0-PI-DXX-000-750, Piping Inspection in Tunnels and Infrequently Accessed Areas, Rev. 0000  
PER 684460, License Renewal NRC Commitment #3: Revise the Buried and Underground
: 0-SI-SXI-067-300.7, System Leakage Test of the Essential Raw Cooling Water System Buried Piping, Rev. 0002 0-TI-DXX-000-915.0, Underground Piping and Tanks Integrity Program, Rev. 0006 G-55, Technical and Programmatic Requirements for the Protective Coating Program for TVA  
Piping and Tanks Inspection Program to Meet License Renewal Requirements
: Nuclear Plants, Rev. 19  
Procedures
: G-94, Piping Installation, Modification and Maintenance, Rev. 2
0-PI-DXX-000-750, Piping Inspection in Tunnels and Infrequently Accessed Areas, Rev. 0000
: NPG-SPP-09.15, Underground Piping and Tanks Integrity Program (UPTI), Rev. 0006  
0-SI-SXI-067-300.7, System Leakage Test of the Essential Raw Cooling Water System Buried
===Other Documents===
Piping, Rev. 0002
: 0901186.000, Structural Integrity Associates, Inc. Baseline Risk Implementation Analysis: Sequoyah Nuclear Power Plant, Rev. 0  
0-TI-DXX-000-915.0, Underground Piping and Tanks Integrity Program, Rev. 0006
: 200931.401, Sequoyah Nuclear Plant Buried Piping Cathodic Protection Design Study, Rev. 0  
G-55, Technical and Programmatic Requirements for the Protective Coating Program for TVA
: Buried Pipe Integrity Program Corrosion Assessment for Buried Piping Systems, dated February 2010
Nuclear Plants, Rev. 19
: CRP-ENG-F-12-002, Assessment of the Underground Piping and Tanks Integrity Program
G-94, Piping Installation, Modification and Maintenance, Rev. 2
: SQN-ENG-S-14-016, Self-Assessment: Readiness for NRC
NPG-SPP-09.15, Underground Piping and Tanks Integrity Program (UPTI), Rev. 0006
: TI 2515/182 Phase 2 Inspection Underground Piping and Tanks Integrity Program Inspection Plan, Rev. 3, dated April 1, 2014
Other Documents
: WO 09-777416-005, Perform UT Examination for Wall Thickness of Excavated Bare Metal on Diesel Fuel Oil Line
0901186.000, Structural Integrity Associates, Inc. Baseline Risk Implementation Analysis:
Sequoyah Nuclear Power Plant, Rev. 0
1200931.401, Sequoyah Nuclear Plant Buried Piping Cathodic Protection Design Study, Rev. 0
Buried Pipe Integrity Program Corrosion Assessment for Buried Piping Systems, dated
February 2010
CRP-ENG-F-12-002, Assessment of the Underground Piping and Tanks Integrity Program
SQN-ENG-S-14-016, Self-Assessment: Readiness for NRC TI 2515/182 Phase 2 Inspection
Underground Piping and Tanks Integrity Program Inspection Plan, Rev. 3, dated April 1, 2014
WO 09-777416-005, Perform UT Examination for Wall Thickness of Excavated Bare Metal on
Diesel Fuel Oil Line
                                                                                    Attachment
}}
}}

Latest revision as of 02:52, 4 November 2019

IR 05000327-14-003 & 05000328-14-003; on 4/1/2014 - 6/30/2014; Sequoyah Nuclear Plant, Units 1 and 2; In-Service Inspection; Radiological Hazard Assessment and Exposure Controls
ML14213A166
Person / Time
Site: Sequoyah  Tennessee Valley Authority icon.png
Issue date: 08/01/2014
From: Bartley J
Reactor Projects Region 2 Branch 6
To: James Shea
Tennessee Valley Authority
References
IR-14-003
Download: ML14213A166 (49)
v • d • e


See also: IR 05000327/2014003

Text

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION II

245 PEACHTREE CENTER AVENUE NE, SUITE 1200

ATLANTA, GEORGIA 30303-1257

August 1, 2014

Mr. Joseph W. Shea

Vice President, Nuclear Licensing

Tennessee Valley Authority

1101 Market Street, LP 3D-C

Chattanooga, TN 37402-2801

SUBJECT: SEQUOYAH NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT

05000327/2014003 AND 05000328/2014003

Dear Mr. Shea:

On June 30, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at

your Sequoyah Nuclear Plant, Units 1 and 2. On July 9, the NRC inspectors discussed the

results of this inspection with Mr. Simmons and other members of your staff. Inspectors

documented the results of this inspection in the enclosed inspection report.

NRC inspectors documented two findings which were determined to be of very low safety

significance (Green) in this report. These findings involved violations of NRC requirements.

If you contest the violation or significance of the NCV, you should provide a response within 30

days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear

Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with

copies to the Regional Administrator, Region II; the Director, Office of Enforcement, U.S.

Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector

at the Sequoyah Nuclear Plant.

If you disagree with a cross-cutting aspect assignment in this report, you should provide a

response within 30 days of the date of this inspection report, with the basis for your

disagreement, to the Regional Administrator, Region II, and the NRC resident inspector at the

Sequoyah Nuclear Plant.

J. Shea 2

In accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections,

Exemptions, Requests for Withholding, 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

NRCs Public Document Room or from the Publicly Available Records (PARS) component of

NRCs Agencywide Documents Access and Management System (ADAMS). ADAMS is

accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public

Electronic Reading Room).

Sincerely,

/RA/

Jonathan H. Bartley, Chief

Reactor Projects Branch 6

Division of Reactor Projects

Docket Nos.: 50-327, 50-328

License Nos.: DPR-77, DPR-79

Enclosure: Inspection Report 050003272014003, 05000328/2014003

w/Attachment: Supplementary Information

cc via ListServ distribution

_________________________ SUNSI REVIEW COMPLETE FORM 665 ATTACHED

OFFICE RII:DRP RII:DRP RII:DRS RII:DRP RII:DRS RII:DRP RII:DRSP

SIGNATURE Via email Via email Via email JHB /RA for/ Via email Via email Via email

NAME GSmith WDeschaine PBraaten CKontz RHamilton WPursley RKellner

DATE 7/31/2014 7/30/2014 7/29/2014 7/31/2014 7/29/2014 7/31/2014 7/30/20148/

E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES/2014 NO

OFFICE RII:DRP RII:DRS RII:DRS RII:DRP RII:DRP

SIGNATURE Via email Via email Via email Via email JHB /RA/

NAME AButcavage ASengupta BCollins JHamman JBartley

DATE 7/31/2014 7/29/2014 7/29/2014 7/31/2014 8/1/20148/

E-MAIL COPY? YES NO YES NO YES NO YES NO YES/2014 NO YES NO YES NO

J. Shea 3

Letter to Joseph W. Shea from Jonathan H. Bartley dated August 1, 2014.

SUBJECT: SEQUOYAH NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT

05000327/2014003 AND 05000328/2014003

Distribution:

D. Gamberoni, RII

L. Douglas, RII

OE Mail

RIDSNRRDIRS

PUBLIC

RidsNrrPMSequoyah Resource

U. S. NUCLEAR REGULATORY COMMISSION

REGION II

Docket Nos.: 50-327, 50-328

License Nos.: DPR-77, DPR-79

Report Nos.: 05000327/2014003, 05000328/2014003

Licensee: Tennessee Valley Authority (TVA)

Facility: Sequoyah Nuclear Plant, Units 1 and 2

Location: Sequoyah Access Road

Soddy-Daisy, TN 37379

Dates: April 1 - June 30, 2014

Inspectors: G .Smith, Senior Resident Inspector

W. Deschaine, Resident Inspector

P. Braaten, Reactor Inspector (1R04)

C. Kontz, Senior Project Engineer (1R05, 1R11, 1R18)

R. Hamilton, Senior Health Physicist (2RS02)

W. Pursley, Health Physicist (2RS01, 2RS03, 2RS04)

R. Kellner, Health Physicist (2RS05)

A. Butcavage, Reactor Inspector (1R08)

A. Sengupta, Reactor Inspector (1R08)

B. Collins, Reactor Inspector (4OA5)

Approved by: Jonathan H. Bartley, Chief

Reactor Projects Branch 6

Division of Reactor Projects

Enclosure

SUMMARY

IR 05000327/2014-003, 05000328/2014-003; 4/1-6/30/2014; Sequoyah Nuclear Plant, Units 1

and 2; In-Service Inspection; Radiological Hazard Assessment and Exposure Controls

The report covered a three-month period of inspection by resident and regional inspectors. Two

findings/violations were identified. The significance of most findings is indicated by their color

(Green, White, Yellow, Red) using 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," Revision 4, dated December 2006.

A. NRC-Identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

Inservice Inspection Requirements was identified for the licensees failure to

perform visual examinations of the control rod drive mechanism (CRDM), American

Society of Mechanical Engineers (ASME) Class 1, seismic plate supports as required

by the ASME Code,Section XI. The licensee entered this issue into their corrective

action program (CAP) as Problem Evaluation Report (PER) 889400. The licensee

developed an operability evaluation and concluded that the supports remained

functional. The licensee also initiated corrective actions to perform the required

visual examinations of the CRDM seismic plate supports before the end of the

current inservice inspection (ISI) interval in April 2016.

The finding was more than minor because it was associated with the protection

against external factors attribute of the mitigating systems cornerstone, and affected

the cornerstone objective to ensure availability, reliability, and capability of systems

that respond to initiating events to prevent undesirable consequence. The inspectors

screened this finding as Green because the finding did not involve the loss or

degradation of equipment or function specifically designed to mitigate a seismic

initiating event. A crosscutting aspect was not assigned to this finding in accordance

with IMC 0612, Appendix B, because the exclusion of the CRDM seismic plate

supports from the scope of the ISI Program occurred outside of the nominal 3-year

period for present performance, and therefore it was not reflective of present

licensee performance. (Section 1R08)

Cornerstone: Occupational Radiation Safety

  • Green: The inspectors identified a Green, self-revealing, NCV of Technical

Specification (TS) 6.12.1, High Radiation Area, for two examples where workers

made entries into High Radiation Areas (HRA) on May 16, 2014, without meeting the

entry requirements specified therein. Specifically, these workers, while performing

decontamination activities and moving materials in the upper reactor containment,

entered a posted HRA: 1) without knowledge of the current radiological conditions in

Enclosure

3

the actual work area, 2) not using a radiological work permit (RWP) approved for

HRA entry, and 3) without wearing the prescribed electronic dosimetry for an HRA.

The licensee entered these events into the Corrective Action Program (CAP) as

Problem Evaluation Reports (PERs) Numbers 886668 and 886160. Immediate

corrective actions included restricting worker access to the Radiologically Controlled

Area (RCA) and issuance of communications to the site and within the Radiation

Protection organization to reinforce roles in RWP adherence and access control.

This finding was more than minor because it is associated with the Occupational

Radiation Safety Cornerstone attribute of Human Performance and adversely affects

the cornerstone objective of ensuring adequate protection of worker health and

safety from exposure to radiation from radioactive material during routine civilian

nuclear reactor operation. The finding was not related to As Low As Reasonably

Achievable planning, nor did it involve an overexposure or substantial potential for

overexposure and the ability to assess dose was not compromised. Therefore, the

finding was determined to be of very low safety significance (Green). This finding

involved the cross-cutting aspect of Human Performance, Avoid Complacency [H.12]

because workers failed to apply appropriate error reduction tools during participation

in the pre-job brief and prior to crossing the HRA boundaries. (2RS1)

B. Licensee-Identified Violations

None.

Enclosure

REPORT DETAILS

Summary of Plant Status:

Unit 1 operated at or near 100 percent rated thermal power (RTP) for the entire inspection

period.

Unit 2 operated at or near 100 percent RTP until April 12, 2014, when the unit entered a power

coast down period. On May 12, with the unit at 76 percent RTP, Unit 2 was shut down for a

refueling outage. Unit 2 returned to 100 percent RTP on June 21, where it operated for the

remainder of the inspection period.

1. REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R04 Equipment Alignment

.1 Partial System Walkdown

a. Inspection Scope

The inspectors performed partial walkdowns of the following two systems to verify the

operability of redundant or diverse trains and components when safety equipment was

inoperable. The inspectors focused on identification of discrepancies that could impact

the function of the system and, therefore, potentially increase risk. The inspectors

reviewed applicable operating procedures, walked down control system components;

and determined whether selected breakers, valves, and support equipment were in the

correct position to support system operation. The inspectors also verified that the

licensee had properly identified and resolved equipment alignment problems that could

cause initiating events or impact the capability of mitigating systems or barriers and

entered them into the corrective action program (CAP). Documents reviewed are listed

in the Attachment. This activity constituted two inspection samples.

  • Spent fuel pool cooling system during Unit 2 core empty period
  • Unit 1 B-train High Head Safety Injection system during A-train planned maintenance

b. Findings

No findings were identified.

Enclosure

5

.2 Complete System Walkdown

a. Inspection Scope

The inspectors performed a complete system walk down of the Unit 2 Main Steam and

support systems to verify proper equipment alignment, to identify any discrepancies that

could impact the function of the system and increase risk, and to verify that the licensee

properly identified and resolved equipment alignment problems that could cause events

or impact the functional capability of the system.

The inspectors reviewed the Updated Final Safety Analysis Report (UFSAR), system

procedures, system drawings, and system design documents to determine the correct

lineup and then examined system components and their configuration to identify any

discrepancies between the existing system equipment lineup and the correct lineup.

During the walkdown, the inspectors reviewed the following:

  • Valves were correctly positioned and did not exhibit leakage that would impact the

functions of any given valve.

  • Electrical power was available as required.
  • Major system components were correctly labeled, lubricated, cooled, ventilated, etc.
  • Hangers and supports were correctly installed and functional.
  • Essential support systems were operational.
  • Ancillary equipment or debris did not interfere with system performance.
  • Valves were locked as required by the locked valve program.
  • Major system components were correctly labeled.
  • Visible cabling appeared to be in good material condition.

In addition, the inspectors reviewed outstanding maintenance work requests and design

issues on the system to determine whether any condition described in those work

requests could adversely impact current system operability. Documents reviewed are

listed in the Attachment. This activity constituted one inspection sample.

b. Findings

No findings were identified.

1R05 Fire Protection

.1 Fire Protection Tours

a. Inspection Scope

The inspectors conducted a tour of the five areas important to safety listed below to

assess the material condition and operational status of fire protection features. The

inspectors evaluated whether: combustibles and ignition sources were controlled in

accordance with the licensees administrative procedures; fire detection and suppression

equipment was available for use; passive fire barriers were maintained in good material

Enclosure

6

condition; and compensatory measures for out-of-service, degraded, or inoperable fire

protection equipment were implemented in accordance with the licensees fire plan.

Documents reviewed are listed in the Attachment. This activity constituted five

inspection samples.

  • Control Building Elevation 669 (Mechanical Equipment Room, 250 VDC Battery and

Battery Board Rooms)

  • Control Building Elevation 685 (Auxiliary Instrument Rooms)
  • Turbine Building Elevation 706
  • Control Building Elevation 706 (Cable Spreading Room)
  • Control Building Elevation 732 (Mechanical Equipment Room and Relay Room)

b. Findings

No findings were identified.

1R06 Flood Protection Measures

Annual Review of Cables Located in Underground Bunkers/Manholes

a. Inspection Scope

The inspectors conducted a review of licensee inspections of safety-related cables

located in underground bunkers/manholes subject to flooding. Specifically, inspectors

reviewed maintenance records of inspections for the previous 12 months to determine if

water was present and, if found, whether it would affect safety-related system operation.

In addition, the inspectors reviewed the licensees corrective action program (CAP) to

ensure that the licensee was identifying underground cabling issues and that they were

properly addressed for resolution. Documents reviewed are listed in the Attachment.

This activity constituted one inspection sample.

b. Findings

No findings were identified.

1R08 Inservice Inspection Activities

a. Inspection Scope

Non-Destructive Examination Activities and Welding Activities: From May 19, 2014,

through May 30, 2014, the inspectors conducted an onsite review of the implementation

of the licensees in-service inspection (ISI) program for monitoring degradation of the

reactor coolant system (RCS), risk-significant piping and components, and containment

systems in Unit 2. The inspectors activities included a review of selected samples of

non-destructive examinations (NDE) to evaluate compliance with the applicable edition

of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel

Enclosure

7

Code (BPVC),Section XI, and to verify that indications and defects (if present) were

appropriately evaluated and dispositioned, in accordance with the requirements of the

ASME Section XI acceptance standards.

The inspectors directly observed the following NDE, mandated by the ASME Code, to

evaluate compliance with the ASME Code Section XI, and Section V requirements, and

if any indications or defects were detected, to evaluate if they were dispositioned in

accordance with the ASME Code, or an NRC-approved alternative requirement.

SGH-4-1, ASME Code Class 2

E-A, Item No E1.30

Inspectors directly observed the calibration of Ultrasonic Test (UT) equipment, and later

reviewed UT examination results for welds associated with a feedwater elbow

attachment to the SG safe end.

Class 2, Augmented Inspection

Class 2, Augmented Inspection

The inspectors reviewed records of the following NDE inspections and methods

mandated by the ASME Code Section XI or augmented inspections, in order to evaluate

compliance with the ASME Code Section XI and Section V requirements, and if any

indications and defects were detected, to evaluate if they were dispositioned in

accordance with the ASME Code or NRC-approved alternative requirements.

  • VT-3, Pipe Support, 2-CVCH-585, ASME Code Class 2
  • VT-3, Pipe Support, 2-CVCH-584, ASME Code Class 2
  • VT-3, Pipe Support, 2-CVCH-586, ASME Code Class 2
  • Penetrant Testing (PT), Integral Attachment Weld, 2-CVCH-006-IA, Code Class 1

The inspectors reviewed the following surface examination records with recordable

indications that were analytically evaluated and accepted for continued service, against

the ASME Code Section XI, or an NRC-approved alternative.

  • PT, Lug to Pipe Weld, 2-SIH-020-IA, ASME Code Class 1

Enclosure

8

No ASME Class 1, 2, or 3 welding activities were in progress during the NRC ISI

inspector site visit. Therefore, the inspectors reviewed the previously completed welding

activity work order (WO), referenced below, in order to evaluate compliance with the

intent of procedures, and the ASME Code. Specifically, the inspectors reviewed the WO

package, the WO VT-2 leakage examination requirements and results.

  • WO No. 112354373, SQN-2-VLV-0012-0817, Valve Replacement, ASME Class 2

Pressurized Water Reactor Vessel Upper Head Penetration Inspection Activities: For

the Unit 2 reactor vessel head, a full bare metal visual (BMV) examination was not

required this outage pursuant to 10 CFR 50.55a. Therefore, no reviews were conducted

for this inspection attribute. A volumetric examination of the Unit 2 vessel upper head

penetration (VUHP) was required this outage. Therefore, inspectors observed and

reviewed a sample of the Unit 2 UT examination results, which included NDE reports for

VUHP Nos. 53, 56, and 60. The inspectors also performed a comparison of the current

UT results to the previous UT examination results for the sample penetrations. These

comparisons were used to determine if the activities, including the disposition of

indications and defects, were conducted in accordance with the requirements of ASME

Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D). In particular, the inspectors

evaluated if the required UT examination scope/coverage was achieved and limitations

(if applicable) were recorded in accordance with the licensee procedures. The licensee

did not identify any relevant indications that needed to be accepted by analysis for

continued service during the UT examination. Additionally, the licensee did not perform

any welding repairs to the vessel head penetrations since the beginning of the last Unit 2

refueling outage; therefore, no NRC review was completed for these inspection

procedure attributes.

Boric Acid Corrosion Control Inspection Activities: The inspectors reviewed the

licensees boric acid corrosion control (BACC) program activities, to ensure

implementation with commitments made in response to NRC Generic Letter 88-05,

Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary, and applicable

industry guidance documents. Specifically, the inspectors performed an onsite record

review of procedures and the results of the licensees containment walkdown inspections

performed during the current spring refueling outage. The inspectors also interviewed

the BACC program owner, conducted an independent walkdown of two levels of

containment, in order to evaluate compliance with the licensees BACC program

requirements, and verify that degraded or non-conforming conditions, such as boric acid

leaks, were properly identified and corrected in accordance with the licensees BACC

and corrective action program (CAP).

The inspectors reviewed the following problem evaluation report (PER), and associated

corrective actions related to evidence of boric acid leakage, to evaluate if the corrective

actions completed were consistent with the requirements of the ASME Code Section XI,

and 10 CFR Part 50, Appendix B, Criterion XVI, and Industry Guidelines.

  • PER 743110, Degraded Non-conforming condition for PDO on RCS leakage and Ice

on Intermediate Deck Doors (IDD), 7/19/13

Enclosure

9

The inspectors reviewed the following engineering evaluations completed for evidence of

boric acid leakage to determine if degraded components were documented in the CAP.

The inspectors also reviewed corrective actions for any degraded components to

determine if they met the applicable requirements of the ASME Code,Section XI, and/or

NRC-approved alternatives.

  • PER 888330, Boric Acid Leakage Evaluation, Reactor Cavity Nozzle Cover Seal

Leaking into Keyway, 6/24/14

  • PER 890230, Evaluation of Boric Acid Corrosion Damage, 2-SNUB-068-RCH30,

6/7/14

  • SR 889942, Determine Available Margins in Pipe Support Attributes, 2-RCH-0028,

5/26/14

Steam Generator Tube Inspection Activities: The inspectors reviewed the eddy current

(EC) examination activities performed in Unit 2 SGs 1, 2, 3, and 4 during the end-of-

cycle 19 refueling outage, to verify compliance with the licensees Technical

Specifications, ASME BPVC Section XI, and Nuclear Energy Institute (NEI) 97-06,

Steam Generator Program Guidelines. The inspectors interviewed licensee personnel

and vendor staff responsible for the SG inspection project, and reviewed documentation

associated with the SG inspections and integrity assessments, as described in this

report section.

The inspectors reviewed the scope of the EC examinations to verify that known and

potential areas of tube degradation were inspected. The inspectors also verified that

inspection scope expansion criteria were implemented based on inspection results, as

directed by the Electric Power Research Institute (EPRI) Pressurized Water Reactor

Steam Generator Examination Guidelines, Revision 7.

The inspectors reviewed documentation for a sample of EC data analysts, EC probes,

and EC testers to verify that personnel and equipment were qualified to detect the

existing and potential degradation mechanisms applicable to Sequoyahs SG tubes, in

accordance with the EPRI Examination Guidelines. This review included a sample of

site-specific Examination Technique Specification Sheets (ETSSs) that were selected

based on plant-specific and industry operating experience, to ensure that their

qualification and site-specific implementation were consistent with Appendix H or I of the

EPRI Examination Guidelines. The selected ETSSs for review consisted of bobbin and

rotating probe techniques that were used to detect wear at the tube interface with

support structures (i.e., tube support plates, anti-vibration bar (AVB), and flow

distribution baffle plate), and wear associated with foreign objects.

The inspectors also reviewed a sample of EC data with a qualified data analyst to

confirm that data analysis was performed in accordance with the applicable ETSSs and

site-specific analysis guidelines. The inspectors verified that the equipment

configuration was consistent with the essential parameters of the applicable technique.

Enclosure

10

The inspectors also verified that recordable indications were detected and sized in

accordance with vendor procedures. As part of the EC data review, the inspectors

verified that the EC indications on each selected tube were consistent with historical

data relative to the number of indications, location, and size. The sample of EC data

selected for review is listed below:

Steam Tube Eddy Current Indication Type

Generator Row/Column Probe

2 R93/C59 Bobbin AVB wear

2 R93/C59 MRPC + point AVB Wear

2 R93/C59 Array AVB Wear

2 R89/C59 Bobbin AVB Wear

2 R89/C59 MRPC + point AVB Wear

4 R93/C47 Bobbin Proximity Signal

4 R93/C47 Array Proximity Signal

2 R5/C101 Bobbin Distorted Support Signal

2 R5/C101 MRPC + point Distorted Support Signal

The inspectors selected a sample of wear degradation mechanisms from the Steam

Generator Degradation Assessment, and verified that the in-situ pressure testing criteria

were determined, in accordance with the EPRI Tube Integrity Guidelines. Additionally,

the inspectors reviewed EC indication reports to determine whether tubes with relevant

indications were appropriately screened for in-situ pressure testing.

The inspectors compared the recent EC examination results with the last Operational

Assessment report for SGs to assess the licensees prediction capability for maximum

tube degradation, and number of tubes with indications. The inspectors verified that the

licensees evaluation was conservative and that current examination results were bound

by the Operational Assessment projections.

The inspectors also compared past examination results discussed in the latest

Degradation Assessment with the recent EC examination results to verify that new

degradation mechanisms, if any, were identified and evaluated before plant startup. The

review of EC examination results included the disposition of potential loose part

indications on the SG secondary side, to verify that corrective actions for evaluating and

retrieving loose parts were consistent with the EPRI Guidelines. The inspectors also

reviewed a sample of primary-to-secondary leakage data for Unit 2 to confirm that

operational leakage in all SGs remained below the action level threshold during the

previous operating cycle.

Based on the review of the final EC examination results for all SGs and interviews with

the licensee, the inspectors confirmed that no EC scope expansion was required, and

none of the SG tubes examined met the criteria for plugging or in-situ pressure testing.

Enclosure

11

Furthermore, the inspectors interviewed licensee staff and reviewed a sample of

secondary side visual inspection results for the SGs 1, 2, 3, and 4 upper bowl areas, to

verify that potential areas of degradation based on site-specific operating experience

were inspected, and appropriate corrective actions were taken to address degradation

indications. This review included the results of Foreign Object Search and Retrieval

(FOSAR) activities in all SGs, and an evaluation for loose parts in the secondary side of

SGs 1, 2, 3, and 4.

Identification and Resolution of Problems: The inspectors reviewed a sample of ISI-

related problems which were identified by the licensee, and entered into the CAP as

PERs. The inspectors reviewed the PERs to confirm that the licensee had appropriately

described the scope of the problem, and had initiated corrective actions. The review

also included the licensees consideration and assessment of operating experience

events applicable to the plant. The inspectors performed this review to ensure

compliance with 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action

requirements. Documents reviewed are listed in the Attachment.

b. Findings

Introduction: An NRC-identified Green NCV of 10 CFR 50.55a (g)(4), Inservice

Inspection Requirements was identified for the licensees failure to perform visual

examinations of the control rod drive mechanism (CRDM), ASME Class 1, seismic plate

supports as required by the ASME Code,Section XI.

Description: The Sequoyah Unit 2 ISI program for the current interval (third interval) was

prepared in accordance with the 2001 Edition of the ASME Section XI Code, with

addenda through 2003. Article IWF-2000 of ASME Section XI, Table IWF-2500-1,

Examination Category Item Number F1.40, requires, in part, periodic VT-3 visual

examinations of supports (other than piping supports) in Class 1 components. The

examinations provide reasonable assurance that the supports can continue to perform

their intended function.

The CRDM assemblies are ASME Class 1 pressure retaining components that contain a

series of seismic plate supports to ensure that the allowable design stress limits for the

CRDM assemblies are not exceeded during a seismic event, which in turn provides

reasonable assurance that the RCS pressure boundary and control rod function is

maintained.

The inspectors identified that the Sequoyah Unit 2 ISI program did not meet the

requirements of ASME Section XI in that the Class 1 CRDM seismic plate supports, and

associated load path components, which meet the examination category F1.40, were not

included in the scope of the program for the first, second, and third ISI intervals. The

inspectors also identified that this issue applied to the Unit 1 ISI Program.

The licensee entered this issue into their CAP as PER 889400. The licensee developed

an operability evaluation and concluded that the supports were operable but non-

conforming. The evaluation considered previous dimensional verifications of the reactor

vessel head lift rig components in the area of the CRDM seismic support plates, and as-

Enclosure

12

found settings of the seismic plates from a modification project WO package associated

with Unit 1 and 2 cables in the seismic plate area of the lift rig. The WO package

included requirements to insert a gap gauge at each seismic plate screw pad gap to

verify the correct gap was present on Unit 1. The results of Unit 1 as-found gap settings

provided reasonable assurance that the as-found gap settings were adequate for Unit 2

based on the similarities in design, operating conditions, and implementation of outage

maintenance activities. The evaluation also considered that no degradation of the lift rig

intervening steel components in the support load path between the seismic plates and lift

rig struts had been reported in previous outages through the CAP. The licensee also

initiated corrective actions to perform the required visual examinations of the CRDM

seismic plate supports before the end of the current ISI interval in April 2016.

Analysis: Failure to perform the required visual examinations of the CRDM seismic

plates and associated load path components, as required by the ASME Section XI Code,

was a performance deficiency (PD). In accordance with Inspection Manual Chapter (IMC) 0612 Appendix B, Issue Screening, the PD was more than minor because it was

associated with the protection against external factors attribute of the mitigating systems

cornerstone, and affected the cornerstone objective to ensure availability, reliability, and

capability of systems that respond to initiating events to prevent undesirable

consequence. Specifically, the licensee failed to perform examinations required to

provide reasonable assurance that the support components can perform their intended

function during design basis seismic events, and therefore maintain the stresses in the

CRDM assembly within the allowable design limits, which in turn provides reasonable

assurance that the RCS pressure boundary and control rod function is maintained. The

inspectors screened this finding as Green in accordance with IMC 0609, Appendix A,

Exhibit 2, Mitigating Systems, because the finding did not involve the loss or degradation

of equipment or function specifically designed to mitigate a seismic initiating event. A

crosscutting aspect was not assigned to this finding in accordance with IMC 0612,

Appendix B, because the exclusion of the CRDM seismic plate supports from the scope

of the ISI Program occurred outside of the nominal 3-year period for present

performance, and therefore it was not reflective of present licensee performance.

Enforcement: Title 10 CFR 50.55a(g)(4), Inservice Inspection Requirements, requires

in part that throughout the service life of a boiling or pressurized water-cooled nuclear

power facility, components (including supports) that are classified as ASME Code Class

1, must meet the requirements, except design and access provisions, and preservice

examination requirements set forth in Section XI of editions and addenda of the ASME

BPVC that become effective subsequent to editions specified in paragraphs (g)(2) and

(g)(3) of this Section, and that are incorporated by reference in paragraph (b) of this

Section, to the extent practical within the limitations of design, geometry, and materials

of construction of the components.Section XI of the ASME BPVC, 2001 Edition with

2003 Addenda, Table IWF-2500-1, Examination Category F-A Supports, requires a VT-3

examination of 100 percent of the ASME Class 1 supports, other than piping supports,

every ISI Interval (examination item F1.40), as modified by Notes 1, 2, 3 and 5 of Table

IWF-2500-1.

Enclosure

13

Contrary to the above, from initial commercial operation until present, the licensee failed

to perform the required VT-3 examination of ASME Class 1 supports, other than piping

supports, (i.e., seismic support plates and associated load path components) on the

CRDM assemblies of Units 1 and 2. The licensee entered the issue into the CAP as

PER 889400. The licensee initiated corrective actions to perform the required VT-3

examinations during the next refueling outage in order to restore compliance with the

10 CFR 50.55a regulations. Because this violation was determined to be of very low

safety significance (i.e., Green), and the licensee entered the issue in the CAP, this

violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC

Enforcement Policy, dated July 9, 2013. This finding will be tracked as NCV 05000327,

328/2014003-01, Failure to Perform Visual Examination of the Unit 1 and Unit 2 CRDM

Seismic Plate Supports.

1R11 Licensed Operator Requalification Program

Quarterly Review

a. Inspection Scope

On June 24, 2014, the inspector observed an evaluated simulator scenario and the

evaluators critique of crew performance. The exercise was performed to provide

practice to the operating crews in longer duration CPE style scenarios. The inspector

observed crew performance in terms of: communications; ability to take timely and

proper actions; prioritizing, interpreting and verifying alarms; correct use and

implementation of procedures, including the alarm response procedures; timely control

board operation and manipulation, including high risk operator actions; oversight and

direction provided by shift manager, including the ability to identify and implement

appropriate Technical Specification (TS) action; and, group dynamics involved in crew

performance. The inspector observed the ability of the licensee to administer the

evaluation and quality of the evaluators critique. The inspector observed scenario

operations for simulator fidelity to verify that it matched actual plant response. Based on

crew performance and scenario administration issues, the inspector also reviewed the

follow-up actions taken to address operator deficiencies and identified administration

issues. Documents reviewed are listed in the Attachment. This activity constituted one

inspection sample.

b. Findings

No findings were identified

.2 Quarterly Review of Licensed Operator Performance

a. Inspection Scope

The inspectors observed and assessed licensed operator performance in the main

control room during periods of heightened activity or risk. The inspectors reviewed

various licensee policies and procedures such as OPDP-1, Conduct of Operations,

NPG-SPP-10.0, Plant Operations, and 0-GO-5, Normal Power Operation. The

Enclosure

14

inspectors utilized activities such as post-maintenance testing, surveillance testing,

unplanned transients, infrequent plant evolutions, plant startups and shutdowns, reactor

power and turbine load changes, and refueling and other outage activities to focus on

the following conduct of operations as appropriate:

  • operator compliance and use of procedures
  • control board manipulations
  • communication between crew members
  • use and interpretation of plant instruments, indications and alarms
  • use of human error prevention techniques
  • documentation of activities, including initials and sign-offs in procedures
  • supervision of activities, including risk and reactivity management
  • pre-job briefs

Specifically, the inspectors observed licensed operator performance during the following

activities:

  • Unit 2 reactor shut down and cool down
  • Unit 2 reactor start up

Documents reviewed are listed in the Attachment. This activity constituted one

inspection sample.

b. Findings

No findings were identified

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed the maintenance activities, issues, and/or systems listed below

to verify the effectiveness of the licensees activities in terms of: appropriate work

practices; identifying and addressing common cause failures; scoping in accordance

with 10 CFR 50.65(b); characterizing reliability issues for performance; trending key

parameters for condition monitoring; charging unavailability for performance;

classification in accordance with 10 CFR 50.65(a)(1) or (a)(2); appropriateness of

performance criteria for structure, system, or components (SSCs) and functions

classified as (a)(2); and appropriateness of goals and corrective actions for SSCs and

functions classified as (a)(1). Documents reviewed are listed in the Attachment. This

activity constituted one inspection sample.

  • Cause Determination Evaluation 2741 associated with failure of flow switch (FS) 2-

FS-74

Enclosure

15

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed the following activities to determine whether appropriate risk

assessments were performed prior to removing equipment from service for

maintenance. The inspectors evaluated whether risk assessments were performed as

required by 10 CFR 50.65(a)(4), and were accurate and complete. When emergent

work was performed, the inspectors reviewed whether plant risk was promptly

reassessed and managed. The inspectors also assessed whether the licensees risk

assessment tool use and risk categories were in accordance with Standard Programs

and Processes Procedure NPG-SPP-07.1, On-Line Work Management, Revision 3,

and Instruction 0-TI-DSM-000-007.1, Risk Assessment Guidelines, Revision 9.

Documents reviewed are listed in the Attachment. The inspectors completed five

samples.

  • Unit 1 Yellow probabilistic safety assessment (PSA) risk associated with 1B Residual

Heat Removal (RHR) pump planned maintenance

  • emergent work due to failure of Individual Rod Position Indication (IRPI) E-5
  • maintenance risk review U2R19 Outage Schedule
  • emergent work due to failure of CS Component Cooling System (CCS) swing pump

hand switch

  • emergent work due to failure of Unit 2 vacuum breaker (2-30-573)

b. Findings

No findings were identified.

1R15 Operability Evaluations

a. Inspection Scope

For the four operability evaluations described in the PERs listed below, the inspectors

evaluated the technical adequacy of the evaluations to ensure that TS operability was

properly justified and the subject component or system remained available, such that no

unrecognized increase in risk occurred. The inspectors compared the operability

evaluations to UFSAR descriptions to determine if the system or components intended

function(s) were adversely impacted. In addition, the inspectors reviewed compensatory

measures implemented to determine whether the compensatory measures worked as

stated and the measures were adequately controlled. The inspectors also reviewed a

sampling of PERs to assess whether the licensee was identifying and correcting any

deficiencies associated with operability evaluations. Documents reviewed are listed in

the Attachment. This activity constituted four inspection samples.

Enclosure

16

  • PER 855557/864224: Operation Decision Making Instruction (ODMI) for Unit 2

Power Range Overpower Rod Stop Deviation Alarms

  • PER 886167: ODMI for Unit 1 Cavity Seal Leakage

24A failure

  • PER 897994: Prompt Determination of Operability (PDO) for Unit 2 Turbine Driven

AFW pump

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing

a. Inspection Scope

The inspectors reviewed the post-maintenance tests associated with the six work orders

(WOs) listed below to assess whether procedures and test activities ensured system

operability and functional capability. The inspectors reviewed the licensees test

procedure to evaluate whether: the procedure adequately tested the safety function(s)

that may have been affected by the maintenance activity, the acceptance criteria in the

procedure were consistent with information in the applicable licensing basis and/or

design basis documents, and the procedure had been properly reviewed and approved.

The inspectors also witnessed the test or reviewed the test data to determine whether

test results adequately demonstrated restoration of the affected safety function(s).

Documents reviewed are listed in the Attachment. This activity constituted six inspection

samples.

  • WO 115806034, Unit 1 Electric Pulse Repair of IRPI Connectors

maintenance and Inspection

  • WO 113875488, Post Maintenance Local Leak Rate Test (as-left) for 2-FCV-63-71,

2-FCV-63-84, & 2-FCV-63-23

b. Findings

No findings were identified.

Enclosure

17

1R20 Refueling and Outage Activities

Unit 2 Refueling Outage Cycle 19

a. Inspection Scope

For the Unit 2 refueling outage that began on May 12, the inspectors evaluated licensee

activities in order to verify that the licensee considered risk in developing outage

schedules, followed risk reduction methods developed to control plant configuration,

developed mitigation strategies for the loss of key safety functions, and adhered to

operating license and TS requirements that ensure defense-in-depth. The inspectors

also walked down portions of Unit 2 not normally accessible during at-power operations

to verify that safety-related and risk-significant SSCs were maintained in an operable

condition. Specifically, between May 12 and June 30, the inspectors performed

inspections and reviews of the following outage activities. This activity constituted one

inspection sample for Refueling Activities.

  • Outage Plan. The inspectors reviewed the outage safety plan and contingency plans

to confirm that the licensee had appropriately considered risk, industry experience,

and previous site-specific problems in developing and implementing a plan that

assured maintenance of defense-in-depth.

  • Reactor Shutdown. The inspectors observed the shutdown in the control room from

the time the reactor was tripped until operators placed it on the RHR system for

decay heat removal to verify that TS cool down restrictions were followed. The

inspectors also toured the lower containment as soon as practicable after reactor

shutdown to observe the general condition of the reactor coolant system (RCS),

emergency core cooling system components, and to look for indications of previously

unidentified leakage inside the polar crane wall.

  • Licensee Control of Outage Activities. On a daily basis, the inspectors attended the

licensee outage turnover meeting, reviewed PERs, and reviewed the defense-in-

depth status sheets to verify that status control was commensurate with the outage

safety plan and in compliance with the applicable TS when taking equipment out of

service. The inspectors further toured the main control room and areas of the plant

daily to ensure that the following key safety functions were maintained in accordance

with the outage safety plan and TS: electrical power, decay heat removal, spent fuel

cooling, inventory control, reactivity control, and containment closure. The

inspectors also observed a tag-out (2-TO-2014-0039, Tag-out of 2B-B Centrifugal

Charging Pump) to verify that the equipment was appropriately configured to safely

support the work and testing. To ensure that RCS level instrumentation was properly

installed and configured to give accurate information, the inspectors reviewed the

installation of the Mansell level monitoring system. Specifically, the inspectors

discussed the system with engineering, walked it down to verify that it was installed

in accordance with procedures and adequately protected from inadvertent damage,

verified that Mansell indication properly overlapped with pressurizer level instruments

during pressurizer drain-down, verified that operators properly set level alarms to

procedurally required set-points, and verified that the system consistently tracked

Enclosure

18

RCS level while lowering to reduced inventory conditions. The inspectors also

observed operators compare the Mansell indications with locally-installed ultrasonic

level indicators during entry into reduced inventory conditions.

  • Refueling Activities. The inspectors observed fuel movement at the spent fuel pool

and at the refueling cavity in order to verify compliance with TS and that each

assembly was properly tracked from core offload to core reload. In order to verify

proper licensee control of foreign material, the inspectors verified that personnel

were properly checked before entering any foreign material exclusion (FME) areas,

reviewed FME procedures, and verified that the licensee followed the procedures.

To ensure that fuel assemblies were loaded in the core locations specified by the

design, the inspectors independently reviewed the recording of the licensees final

core verification.

  • Reduced Inventory and Mid-Loop Conditions. Prior to the outage, the inspectors

reviewed the licensees commitments to Generic Letter 88-17. Before entering

reduced inventory conditions the inspectors verified that these commitments were in

place, that plant configuration was in accordance with those commitments, and that

distractions from unexpected conditions or emergent work did not affect operator

ability to maintain the required reactor vessel level. The inspectors verified that

licensee procedures for closing the containment upon a loss of decay heat removal

were in effect, that operators were aware of how to implement the procedures, and

that other personnel were available to close containment penetrations, if needed. In

order to reduce outage risk, the licensee elected to not put the plant into mid-loop

conditions during this particular refueling outage.

  • Heatup and Startup Activities. The inspectors toured the containment prior to reactor

startup to verify that debris that could affect the performance of the containment

sump had not been left in the containment. The inspectors reviewed the licensees

mode-change checklists to verify that appropriate prerequisites were met prior to

changing TS modes. Prior to plant startup, the inspectors performed a detailed tour

of containment to ensure no debris existed that could affect containment sump

performance given a design basis accident. The inspectors also inspected the

primary system in containment during Mode 3 with the plant at normal operating

pressure and temperature in order to verify the leak tightness of the RCS.

b. Findings

No findings were identified.

1R22 Surveillance Testing

a. Inspection Scope

For the 11 surveillance tests identified below, the inspectors assessed whether the

SSCs involved in these tests satisfied the requirements described in the TS surveillance

requirements, the UFSAR, applicable licensee procedures, and whether the tests

demonstrated that the SSCs were capable of performing their intended safety functions.

Enclosure

19

This was accomplished by witnessing testing and/or reviewing the test data. Documents

reviewed are listed in the Attachment. This activity constituted 11 inspection samples.

In-Service Tests:

  • 2-SI-SXP-063-202.0, Safety Injection Pumps 2A-A and 2B-B Comprehensive

Performance and Check Valve Test, Revision 5

Routine Surveillance Tests:

  • 0-SI-OPS-092-078.0 Power Range Neutron Flux Channel Calibration by Heat

Balance Comparison, Revision 23

  • 2-SI-OPS-082-026.B, Loss of Offsite Power with Safety Injection - D/G 2B-B Test,

Revision 43

  • 2-SI-OPS-088-001.0, Phase A Isolation Test, Revision 20
  • 2-SI-OPS-082-026.A, Loss of Offsite Power with Safety Injection - D/G 2A-A Test,

Revision 47

  • 2-SI-OPS-000-009.0, Actuation of ECCS and Boron Injection Flow path valves via

Safety Injection Signal, Revision 9

Ice Condenser Surveillance Test:

  • 0-SI-MIN-061-107.0, Ice Condenser Floor Drains, Revision 2
  • 0-SI-MIN-061-109.0, Ice Condenser Intermediate and Lower Inlet Doors and Vent

Curtains, Revision 5

Containment Isolation Valve (CIV) Surveillance Tests:

  • 0-SI-SLT-062-258.1, Containment Isolation Valve Local Leak Rate Test Chemical

and Volume Control System, Revision 11

  • 2-SI-OPS-088-003.0, Phase B Containment Isolation Test, Revision 10

b. Findings

No findings were identified.

Cornerstone: Emergency Preparedness

1EP6 Drill Evaluation

a. Inspection Scope

Resident inspectors evaluated the conduct of a routine licensee emergency drill on

April 1, 2014, to identify any weaknesses and deficiencies in classification, notification,

and protective action recommendation (PAR) development activities. This drill involved

beyond design basis events and utilized the licensees severe accident mitigation

Enclosure

20

guidelines (SAMG). The inspectors evaluated emergency response operations in the

simulated control room, as well as the technical support center, to verify that event

classification and notifications were performed in accordance with EPIP-1, Emergency

Plan Classification Matrix, Revision 51. The inspectors verified that the licensee properly

utilized the SAMGs. The inspectors also attended the licensee critique of the drill to

compare any inspector observed weakness with those identified by the licensee in order

to verify whether the licensee was properly identifying deficiencies. This activity

constituted one inspection sample.

b. Findings

No findings of significance were identified.

2. RADIATION SAFETY (RS)

Cornerstones: Occupational Radiation Safety and Public Radiation Safety

2RS1 Radiological Hazard Assessment and Exposure Controls

a. Inspection Scope

Hazard Assessment and Instructions to Workers: During facility tours, the inspectors

directly observed labeled radioactive material and postings for radiation areas and High

Radiation Areas (HRAs) established within the Radiologically Controlled Area (RCA) of

the Unit 2 (U2) upper and lower containments, Auxiliary Building and Dry Active Waste

(DAW) Storage Facility. The inspectors independently measured radiation dose rates or

directly observed conduct of licensee radiation surveys for selected RCA areas,

including the Independent Spent Fuel Storage Installation (ISFSI). The inspectors

reviewed and verified survey records for several plant areas including surveys for alpha

emitters, airborne radioactivity, and gamma surveys with a range of dose rate gradients.

The inspectors reviewed several radiation work permit (RWP) details to assess

communication of radiological control requirements and current radiological conditions to

workers. The inspectors reviewed selected Electronic Dosimeter (ED) dose and dose

rate alarms, to verify workers properly responded to the alarms and that the licensees

review of the events was appropriate. The inspectors observed jobs in radiologically

risk-significant areas including HRAs and areas with, or with the potential for airborne

activity.

Contamination and Radioactive Material Control: The inspectors observed the release

of potentially contaminated items from the RCA and from contaminated areas (i.e., U2

containment). The inspectors also reviewed the procedural requirements for, and

equipment used to perform, the radiation surveys for release. During plant walk downs,

the inspectors evaluated radioactive material storage areas and containers, including

satellite RCAs and yard areas, assessing material condition, posting/labeling, and

control of materials/areas. In addition, the inspectors reviewed the sealed source

inventory and verified labeling, storage conditions, and leak testing of selected sources.

Enclosure

21

Radiological Hazards Control and Work Coverage: The inspectors evaluated licensee

performance in controlling worker access to radiologically significant areas and

monitoring jobs in-progress during the week of the onsite inspection. The inspectors

also reviewed the procedural guidance for multi and extremity badging. For HRA tasks

involving significant dose rate gradients, the inspectors evaluated the use and placement

of whole body and extremity dosimetry to monitor worker exposure. The inspectors

reviewed RWPs for use in airborne areas, ensuring the prescribed controls were

appropriate for the conditions as identified in radiological surveys and air samples. ED

alarm set points and worker stay times were evaluated against area radiation survey

results for containment and auxiliary building activities.

Risk Significant High Radiation Areas and Very High Radiation Area Controls: The

inspectors evaluated access barrier effectiveness for selected Locked High Radiation

Area (LHRA) and Very High Radiation Area (VHRA) locations. Changes to procedural

guidance for LHRA and VHRA controls were discussed with Radiation Protection (RP)

supervisors. During plant walk downs of the U2 Containment and Auxiliary Building, the

inspectors verified the posting/locking of LHRA/VHRA areas. Established radiological

controls (including airborne controls) were evaluated for selected tasks including work in

auxiliary building HRAs, and radiological waste processing and storage. In addition,

licensee controls for areas where dose rates could change significantly as a result of

plant shutdown and refueling operations were reviewed and discussed.

Radiation Worker Performance and RP Technician Proficiency: The inspectors

observed radiation worker performance through direct observation. Jobs observed

included routine waste packaging activities in the auxiliary building and routine survey

activities in the Auxiliary Building and Upper and Lower Containments in high radiation

and contaminated areas. The inspectors also observed health physics technicians

(HPTs) providing pre-job/RWP briefings, releasing material from the RCA, and providing

field coverage of jobs. Occupational workers adherence to selected RWPs and HPT

proficiency in providing job coverage were evaluated through direct observations and

interviews with licensee staff. ED alarm set points and worker stay times were evaluated

against area radiation survey results for reviewed RWPs.

Problem Identification and Resolution: PERs associated with radiological hazard

assessment and control were reviewed and assessed. The inspectors evaluated the

licensees ability to identify, characterize, prioritize, and resolve the identified issues in

accordance with procedure NPG-SPP-22-300, Corrective Action Program, (CAP)

Revision (Rev.) 1. The inspectors also evaluated the scope of the licensees internal

audit program and reviewed recent assessment results.

RP activities were evaluated against the requirements of Updated Final Safety Analysis

Report (UFSAR) Section 12; Technical Specifications (TS) Sections 6.12; 10 CFR Parts 19 and 20; and approved licensee procedures. Licensee programs for monitoring

materials and personnel released from the RCA were evaluated against 10 CFR Part 20

and IE Circular 81-07, Control of Radioactively Contaminated Material. Documents

reviewed are listed in the Attachment.

Enclosure

22

b. Findings

Introduction: The inspectors identified a Green, self-revealing, Non-cited Violation

(NCV) of TS 6.12.1, High Radiation Area, for two examples of individuals entering an

HRA without meeting the entry requirements as specified in TS 6.12.1.b and e.

Description: On May 16, 2014, with the U2 reactor shutdown for refueling, a contract

employee who was staging equipment and two contract decontamination technicians,

working on separate jobs in the upper reactor containment, entered the same posted

HRA near the reactor cavity. One of the decontamination technicians and the contractor

staging equipment received dose rate alarms shortly after crossing the HRA boundary.

Upon receiving the alarms, both individuals exited the area and contacted health physics

(HP) as required. The two decontamination technicians were on RWP Number (No.)

210061 with a dose setpoint of 31 mrem and dose rate setpoint of 91 milli-rem per hour

(mrem/hr). The decontamination workers ED indicated a maximum dose rate of 97

mrem/hr. The worker staging equipment was on RWP No. 240051 with a dose setpoint

of 21 mrem and dose rate setpoint of 81 mrem/hr. That workers ED indicated a

maximum dose rate of 133 mrem/hr. Accessible general area dose rates based on

surveys in the area near the time of the event were as high as 160 mrem/hr at 30

centimeters (cm).

In both cases the workers had only been briefed for entry into Radiation Areas in the

upper reactor containment and that expected dose rates in this area were 3-10 mrem/hr.

They were not wearing the prescribed alarming dosimetry for an HRA entry, were not on

an RWP that allowed HRA entry, and were not knowledgeable of the actual dose rates in

the area. The licensee entered these events into their CAP (PERs 886668 and 886160).

Immediate corrective actions included restricting worker access to the RCA and

issuance of communications to the site and within the RP organization to reinforce roles

in RWP adherence and access control.

Analysis: The inspectors determined that entry into a HRA without meeting the entry

requirements specified in TS 6.12.1 was a performance deficiency. This finding is more

than minor because it is associated with the Occupational Radiation Safety Cornerstone

attribute of Human Performance and adversely affects the cornerstone objective of

ensuring adequate protection of worker health and safety from exposure to radiation

from radioactive material during routine civilian nuclear reactor operation. Workers

permitted entry into HRAs with inadequate knowledge of actual radiological conditions

could receive unintended occupational exposures. The finding was evaluated using the

Occupational Radiation Safety Significance Determination Process (SDP). The finding

was not related to ALARA planning, nor did it involve an overexposure or substantial

potential for overexposure, and the ability to assess dose was not compromised.

Therefore, the inspectors determined the finding to be of very low safety significance

(Green). The inspectors noted that the workers responded properly to the ED dose rate

alarms thereby limiting their potential for unintended exposure. This finding involved the

cross-cutting aspect of Human Performance, Avoid Complacency [H.12] because

workers failed to apply appropriate error reduction tools while participating in pre-job

briefs and prior to crossing the HRA boundaries.

Enclosure

23

Enforcement: TS 6.12.1, High Radiation Area, requires in part, entries into HRAs be

controlled by means of an RWP, associated radiation surveys, and other appropriate

radiation protection equipment and measures and except for individuals qualified in RP

procedures or escorted by such individuals, entry into such areas shall only be made

after dose rates in the area have been determined and entry personnel are made

knowledgeable of them. Contrary to the above, on May 16, 2014, workers entered a

HRA using an RWP that did not allow HRA access, without using the proper alarming

dosimetry, and without knowledge of the actual dose rates in the area. Because this

violation was of very low safety significance and it was entered into the licensees CAP

(PERs 886668 and 886160), this violation is being treated as an NCV, consistent with

the Enforcement Policy: NCV 05000327/328, 2014003-02, Failure to Comply with Entry

requirements to a HRA.

2RS2 Occupational ALARA Planning and Controls

a. Inspection Scope

Work Planning and Exposure Tracking: The inspectors reviewed work activities and

their collective exposure estimates associated with the previous Unit 1 (U1) refueling

outage, as well as the current U2 refueling outage 19 (U2R19). The U1 refueling outage

19 (U1R19) and U2R19 ALARA planning packages (ALARA Plans) were reviewed for

the following high collective exposure tasks: Refueling operations, Mechanical

Maintenance, Plant Services, RP and Modifications. For the selected tasks, the

inspectors reviewed the assumptions and basis for the dose rate and man-hour

estimates. The inspectors discussed with ALARA staff the means by which wrench-

hours were derived from the work order hours provided by craft supervision to ALARA

staff. The inspectors verified the licensee had established several means to track and

trend doses for ongoing work activities. The inspectors evaluated the incorporation of

exposure reduction initiatives and operating experience, including historical post-job

reviews, into RWP requirements. Collective dose data for selected tasks were

compared with established dose estimates and evaluated against procedural criteria

(trigger points) for additional ALARA review. Where applicable, changes to established

estimates were discussed with ALARA planners and evaluated against work scope

changes or unanticipated elevated dose rate. The inspectors discussed the operation of

the Station ALARA Committee with the Site Vice President, the RP Manager and the

ALARA Health Physicist. For ALARA Plans from U1R19, the inspectors compared the

results achieved in terms of actual dose versus (vs.) planned dose and actual hours vs.

estimated hours, reviewed in-progress and post-job ALARA reviews, and discussed the

job planning, performance, and reviews with ALARA staff. For ALARA Plans associated

with U2R19, the inspectors reviewed dose-to-date on select jobs, comparing estimates

with actuals, and observed development of selected in-progress reviews.

Source Term Reduction and Control: The inspectors reviewed the collective exposure

three-year rolling average (TYRA) from 2011 - 2013 and reviewed historical outage

collective exposure trends. Through interviews with licensee staff and document review,

the inspectors assessed the licensees current activities related to source term reduction,

including elevated zinc injection on U2, on-line chemistry using pH 7.4 to minimize

corrosion product transport, extended reactor coolant pump run time to allow better

Enclosure

24

cleanup during shutdown, ultrasonic fuel cleaning, and response to fuel defects during

previous operating cycles. The inspectors discussed the unexpectedly high activity of

shutdown crud burst and changes expected in the short and long term relative

abundances of Cobalt-58 and Cobalt-60 that would result from the change in the steam

generator tube alloys and increasing the number of steam generator tubes by about a

third. The dose implications of the various cobalt reduction activities coupled to the

change in tube alloys for the next few outages was also discussed

Radiation Worker Performance: Radiation worker performance was also observed and

evaluated as part of Inspection Procedure 71124.01 and is documented in section 2RS1.

While observing job tasks, the inspectors evaluated the use of remote technologies to

reduce dose including teledosimetry and remote visual monitoring. Jobs observed were

associated with the refueling and maintenance outage.

Problem Identification & Resolution: Licensee CAP documents associated with ALARA

planning and controls were reviewed and assessed. This included a review of selected

Action Requests (PERs), self-assessments, and audits. The inspectors evaluated the

licensees ability to identify, characterize, prioritize, and resolve the identified issues in

accordance with procedure NPG-SPP-22.300, Corrective Action Program, Rev. 1. The

inspectors also evaluated the scope and frequency of the licensees self-assessment

program and reviewed recent assessment results.

ALARA program activities were evaluated against the requirements of UFSAR Section

12, Radiation Protection; TS Section 6.8, Procedures and Programs; 10 CFR Part 20;

and approved licensee procedures. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

2RS3 In-Plant Airborne Radioactivity Control and Mitigation

a. Inspection Scope

Engineering Controls: The inspectors reviewed the use of temporary and permanent

engineering controls to mitigate airborne radioactivity during U2R19 for steam generator

(S/G) and U2 Thimble eddy current testing and the DAW Storage Building. The use of

the U2 Containment Purge to minimize airborne concentrations in containment during

refuel activities was discussed with licensee personnel. In addition, inspectors observed

the placement and use of high efficiency particulate air negative pressure units, and air

sampling equipment during observations of jobs in-progress.

Use of Respiratory Protection Devices & Self-Contained Breathing Apparatus for

Emergency Use: Inspectors reviewed the use of respiratory protection devices to limit

the intake of radioactive material, including devices used for routine tasks and devices

stored for use in emergency situations. Inspectors observed the physical condition of

Self-Contained Breathing Apparatus (SCBA) units, negative pressure respirators

(NPR)s, powered air purifying respirators and device components staged for routine and

Enclosure

25

emergency use throughout the plant. SCBA bottle air pressure, the number of units, and

the number of spare masks and air bottles available was also evaluated by inspectors.

The inspectors reviewed maintenance records for selected SCBA units for the past year

and evaluated SCBA and NPR compliance with National Institute for Occupational

Safety and Health certification requirements. The inspectors also reviewed records of

Grade D (or better) air quality testing for supplied-air devices and SCBA bottles. In

addition, the inspectors walked-down the compressor used for filling SCBA bottles. The

inspectors reviewed the status and surveillance records of SCBAs staged for in-plant

use during emergencies through review of records and walk-down of SCBA staged in

the control room and selected locations.

The inspectors verified the licensee had procedures in place to ensure that the use of

respiratory protection equipment was ALARA when engineering controls were not

practicable. Control room operators and fire brigade were interviewed on the use of the

devices including SCBA bottle change-out and use of corrective lens inserts. Respirator

qualification records were reviewed and cross checked for several control room

operators. In addition, qualifications for individuals responsible for testing and repairing

SCBA vital components were evaluated through review of training records. Selected

maintenance records for SCBA units and air cylinder hydrostatic testing documentation

were reviewed.

The inspectors verified that the licensee has procedural requirements in place for

evaluating air samples for the presence of alpha emitters and reviewed airborne

radioactivity and contamination survey records for selected plant areas to ensure air

samples are screened and evaluated per the procedure requirements.

The inspectors walked-down the respirator issue and storage locations and verified that

the equipment was appropriately stored and maintained. Records of monthly and

quarterly inventory and inspection of the equipment were also reviewed by the

inspectors. The inspectors discussed the process for issuing respirators, and verified

that selected individuals qualified for respirator and/or SCBA use had completed the

required training, fit-test, and medical evaluation.

Problem Identification and Resolution: Licensee CAP documents associated with the

control and mitigation of in-plant radioactivity were reviewed and assessed. This

included review of selected PERs related to use of respiratory protection devices

including SCBA. The inspectors evaluated the licensees ability to identify, characterize,

prioritize, and resolve the identified issues in accordance with procedure NPG-SPP-22-

300, Corrective Action Program, Rev.1. The inspectors also evaluated the scope of the

licensees internal audit program and reviewed recent assessment results.

RP activities were evaluated against the requirements UFSAR Section 12; 10 CFR Parts

19 and 20; and approved licensee procedures. Documents and records reviewed are

listed in the Attachment.

b. Findings

No findings were identified.

Enclosure

26

2RS4 Occupational Dose Assessment

a. Inspection Scope

External Dosimetry: The inspectors reviewed National Voluntary Laboratory

Accreditation Program certification data and discussed program guidance for storage,

processing, and evaluation of results for active and passive personnel dosimeters

currently in use. Comparisons between ED and thermo-luminescent dosimeter data

were discussed in detail. The inspectors reviewed ED alarm logs and reviewed

licensees dosimeter incident reports and assessment actions for selected alarm events.

Internal Dosimetry: Program guidance and assessment results for internally deposited

radionuclides were reviewed. The inspectors reviewed selected Whole Body Count (in

vivo) analyses from September 2012 to May 2014 as well as in-vitro assessments of

tritium exposures to workers entering Unit 2 containment at power during this period.

The licensees methods used in these assessments as well as the programs for

collection and analysis of special bioassay samples were discussed with licensee staff.

Special Dosimetric Situations: The inspectors evaluated the licensees use of multi-

badging, extremity dosimetry, and dosimeter relocation within non-uniform dose rate

fields and reviewed assessments for U2R19 for S/G maintenance workers. Worker

monitoring in neutron areas was discussed with licensee staff. The inspectors also

reviewed records of monitoring for declared pregnant workers from September 2012 to

May 2014 and discussed monitoring guidance with dosimetry staff. In addition, methods

for shallow dose assessments were reviewed and discussed.

Problem Identification and Resolution: The inspectors reviewed and discussed selected

CAP documents associated with occupational dose assessment. The inspectors

evaluated the licensees ability to identify and resolve the issues in accordance with

procedure NPG-SPP-22-300, Corrective Action Program, Rev.1. The inspectors also

discussed the scope of the licensees internal audit program and reviewed recent

assessment results.

Occupational dose assessment activities were evaluated against the requirements of

UFSAR Section 12; TS Section 6; 10 CFR Parts 19 and 20; and approved licensee

procedures. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

2RS5 Radiation Monitoring Instrumentation

a. Inspection Scope

Radiation Monitoring Instrumentation: During walk-downs of the auxiliary building and

the RCA exit point, the inspectors observed installed radiation detection equipment.

These included area radiation monitors (ARMs), liquid and gaseous effluent monitors,

Enclosure

27

personnel contamination monitors (PCMs), small article monitors (SAMs), and portal

monitors (PMs). The inspectors observed the physical location of the components and

noted their material condition.

In addition to equipment walk-downs, the inspectors reviewed source checks of various

portable and fixed detection instruments, including ion chambers, teletectors, PCMs,

SAMs, PMs, and an iSOLO alpha/beta counting system. The inspectors reviewed

calibration records and evaluated alarm set-point values for PCMs, PMs, effluent

monitors, an ARM, and a SAM. This included a sampling of instruments used for post-

accident monitoring such as a containment high-range radiation monitor and effluent

monitors for noble gas and iodine. The radioactive source used to calibrate an effluent

monitor was evaluated for traceability to national standards. Calibration stickers on

portable survey instruments were noted during inspection of the storage area for ready-

to-use equipment. The most recent 10 CFR Part 61 analysis for DAW was reviewed to

determine if calibration and check sources are representative of the plant source term.

The inspectors also reviewed count room calibration records for a gamma spectroscopy

germanium detector and a liquid scintillation detector.

Effectiveness and reliability of selected radiation detection instruments were reviewed

against details documented in the following: 10 CFR Part 20; NUREG-0737,

Clarification of TMI Action Plan Requirements; UFSAR Chapters 11 and 12; and

applicable licensee procedures.

Problem Identification and Resolution: The inspectors reviewed selected PER reports in

the area of radiological instrumentation. The inspectors evaluated the licensees ability

to identify and resolve the issues in accordance with procedure NPG-SPP-22.300,

Corrective Action Program, Rev. 1. Documents and records reviewed are listed in the

Attachment.

b. Findings

No findings were identified.

4. OTHER ACTIVITIES

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency

Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Security

4OA1 Performance Indicator (PI) Verification

a. Inspection Scope

The inspectors sampled licensee submittals for the five PIs listed below for the period

from January 2013 through March 2014 for both Unit 1 and Unit 2. Definitions and

guidance contained in Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment

Indicator Guideline, Revision 6, were used to determine the reporting basis for each data

element in order to verify the accuracy of the PI data reported during that period.

Enclosure

28

Cornerstone: Initiating Events

  • Unplanned Scrams per 7000 Critical Hours
  • Unplanned Scrams with Complications

The inspectors reviewed selected Licensee Event Reports (LERs) and portions of

operator logs to verify whether the licensee had accurately identified the number of

scrams and unplanned power changes that occurred during the previous four quarters

for both units. The inspectors also reviewed the accuracy of the number of critical hours

reported and the licensees basis for addressing the criteria for complications for each of

the reported scrams. Documents reviewed are listed in the Attachment.

Cornerstone: Occupational Radiation

  • Occupational Exposure Control Effectiveness

The inspectors reviewed PI data collected from November 2013 through May 2014, for

the Occupational Exposure Control Effectiveness PI. For the reviewed period, the

inspectors assessed PER records to determine whether HRA, VHRA or unplanned

exposures, resulting in TS or 10 CFR 20 non-conformances, had occurred during the

review period. The inspectors reviewed RCA exit transactions with exposures in excess

of 100 milli-rem in order to determine compliance with the requirements of the RWP.

The reviewed data were assessed against guidance contained in Nuclear Energy

Institute (NEI) 99-02, "Regulatory Assessment Indicator Guideline," Rev. 6.

Cornerstone: Public Radiation Safety

  • Radiological Control Effluent Release Occurrences

The inspectors reviewed the Radiological Control Effluent Release Occurrences PI

results for the Public Radiation Safety Cornerstone from November 2013, through May

2014. For the assessment period, the inspectors reviewed cumulative and projected

doses to the public and PER documents related to Radiological Effluent Technical

Specifications/Offsite Dose Calculation Manual issues.

b. Findings

No findings were identified.

Enclosure

29

4OA2 Problem Identification and Resolution

.1 Daily Review

a. Inspection Scope

As required by Inspection Procedure 71152, Identification and Resolution of Problems,

and in order to help identify repetitive equipment failures or specific human performance

issues for follow-up, the inspectors performed a daily screening of items entered into the

licensees CAP. This was accomplished by reviewing the description of each new PER

and attending daily management review committee meetings.

b. Findings and Observations

No findings were identified.

.2 Selected Issue Follow-up: Pressurizer Thermal Limit Exceeded

a. Inspection Scope

The inspectors performed an in-depth review of PER 809100. On November 9, 2013,

during a plant heat-up on Unit 1, the pressurizer thermal limits were exceeded on two

separate occurrences, contrary to the Technical Requirements Manual (TRM) Section

3.9.2. Westinghouse performed an analysis of the event and determined that the

existing pressurizer design basis analysis bounded this event and there was no increase

in the limiting primary stress or the primary-plus-secondary stress range from this event.

As a result of this event, the licensee developed an Apparent Cause Analysis (ACE).

The inspectors reviewed the actions taken to determine if the licensee had adequately

addressed the following attributes.

  • Complete, accurate and timely identification of the problem
  • Evaluation and disposition of operability and reportability issues
  • Consideration of previous failures, extent of condition, generic or common cause

implications

  • Prioritization and resolution of the issue commensurate with safety significance
  • Identification of the root cause and contributing causes of the problem
  • Identification and implementation of corrective actions commensurate with the safety

significance of the issue

b. Findings

There were no findings identified during this review. The inspectors determined that the

ACE was probing and involved an extent of condition review, a safety culture analysis,

and operating experience review. The ACE also brought to light that the crew that

performed the pressurizer heat up did not realize the limit violation. This fact was

actually noted by the night shift crew when reviewing data. The ACE also analyzed a

near miss that occurred on November 13, 2013. In this instance, the plant was being

Enclosure

30

returned to cold iron conditions due to a pressurizer power operated relief failure and the

operators were challenged in maintaining cool-down limits of the pressurizer. Although

no limits were exceeded the November 13 incident, the ACE noted several weaknesses

in the evolution. The ACE ultimately led to the development of several corrective

actions; including procedural changes designed strengthen the operators awareness of

pressurizer pressure control, development of a vendor-performed stress analysis of the

event, and addition of the event to the operations training program in order to share

information with other crews.

.3 Semi-Annual Trend Review

a. Inspection Scope

As required by Inspection Procedure 71152, the inspectors performed a semi-annual

review of the licensees corrective action program and associated documents to identify

trends that could indicate the existence of a more significant safety issue. The

inspectors review was focused on repetitive equipment issues, but also included

licensee trending efforts and licensee human performance results. The inspectors

review nominally considered the twelve-month period of July 2013 through June 2014,

although some examples expanded beyond those dates when the scope of the trend

warranted. Specifically, the inspectors considered the results of daily inspector

screening discussed in Section 4OA2.1 and reviewed licensee trend reports for the

period in order to determine the existence of any adverse trends that the licensee may

not have previously identified. This activity constituted one inspection sample for Semi-

annual Trend Review.

b. Findings and Observations

No findings were identified. The inspectors noted a negative trend regarding human

performance errors. During the daily reviews, the inspectors noted an increase in

human error events. The inspectors then performed a more detailed review of the trend

under the semi-annual trend review required by IP 71152. The inspectors concluded

there were at least eleven of these events that occurred in the last three months. The

inspectors noted this was more than the typical amount of error-related incidents

observed during a quarter. The below abbreviated list of PERs involved several human

performance related and mis-positioning events as well as procedural non-compliance.

  • PER 868301, EDG 1B and 2B Fan Switch in Incorrect Position, (April 4, 2014)
  • PER 876825, Vent Valve Found in Wrong Position, (April 25)
  • PER 877256, A EGTS damper switch found in incorrect Position, (April 27)
  • PER 878321, B Train Purge Aligned with A Train Radiation Monitor, (April 30)
  • PER 878588, Missing Locking Mechanism on Charging Valve (April 30)
  • PER 882745, Switch Error Alignment of Inverter during Testing, (May 9)
  • PER 884002, Boric Acid Valve Found in Wrong Position, (May 13)
  • PER 884012, Danger-Tagged Switch Found in Wrong Position, (May 13)
  • PER 885856, Incorrect Pressurizer Safety Valve Removed, (May 16)

Enclosure

31

  • PER 886066, Missed QC Hold Point, (May 17)
  • PER 886765, RHR Valves Found in Wrong Position, (May 19)

The residents discussed this negative human performance trend with site management.

Most of the errors involved some form of procedural non-compliance. The licensee

concurred with the observation and noted that they had also concurrently and

independently (of the NRC resident staff) identified the same trend. This was

documented in PER 884559 and generated on May 14. Immediate corrective actions to

these errors included stand-downs emphasizing procedural compliance with the craft

personnel and site-wide communications to remind staff to use error reduction tools

when performing high risk activities. The inspectors noted that the licensee was

aggressively dealing with these human performance deficiencies and a reasonable

assurance exists that the trend can be reversed. Although these issues should be

corrected, they constitute violations of minor significance that are not subject to

enforcement action in accordance with Section 2 of the Enforcement Policy.

4OA5 OTHER ACTIVITIES

.1 (Closed) Temporary Instruction 2515/182 - Review of the Industry Initiative to Control

Degradation of Underground Piping and Tanks

a. Inspection Scope

The inspectors conducted a review of records and procedures related to the licensees

program for buried piping and underground piping and tanks in accordance with

Phase II of temporary instruction (TI) 2515/182 to confirm that the licensees program

contained attributes consistent with Sections 3.3.A and 3.3.B of Nuclear Energy

Institute (NEI) 09-14, Guideline for the Management of Buried Piping Integrity,

Revision 3, and to confirm that these attributes were scheduled and/or completed by

the NEI 09-14 deadlines. The inspectors interviewed licensee staff responsible for the

buried piping program and reviewed program related activities to determine if the

program attributes were accomplished in a manner which reflected acceptable

practices in program management.

The licensees buried piping and underground piping and tanks program was inspected

in accordance with paragraph 03.02.a of the TI and it was confirmed that activities,

which correspond to completion dates specified in the program which have passed

since the Phase 1 inspection was conducted, have been completed. The licensees

buried piping and underground piping and tanks program was inspected in accordance

with paragraph 03.02.b of the TI and responses to specific questions found in

http://www.nrc.gov/reactors/operating/ops-experience/buried-pipe-ti-phase-2-insp-req-

2011-11-16.pdf were submitted to the NRC headquarters staff. Additionally, the

inspectors reviewed the licensees risk ranking process and implementation of the

inspection plan using the guidance of paragraph 03.04 and 03.05 of the TI.

Enclosure

32

b. Findings

No findings were identified. Based upon the scope of the review described above,

Phase 2 of TI-2515/182 was completed.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On July 9, 2014, the resident inspectors presented the inspection results to

Mr. Simmons and other members of his staff, who acknowledged the findings. No

proprietary information was discussed.

ATTACHMENT: SUPPLEMENTAL INFORMATION

Enclosure

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

J. Carlin, Site Vice President

A. Day, Chemistry Manager

D. Erb, Work Control Manager

B. Gann, Dosimetry/Instruments Supervisor

M. Henderson, ISI Program Engineer

J. Johnson, Program Manager Licensing

T. Johnston, Radiation Protection Support Manager

K. Loomis, Site Program Owner

T. Marshall, Operations Manager

M. McBrearty, Licensing Manager

T. Noe, Director Safety and Licensing

W. Pierce, Site Engineering Director

P. Pratt, Maintenance Manager

R. Rice, Radiation Protection Manager

J. Rolph, Radiation Protection Technical Support Superintendent

P. Simmons, Plant Manager

K. Smith, Director of Training

C. Summers, Health Physicist-ALARA

NRC personnel

S. Lingam, Project Manager, Office of Nuclear Reactor Regulation

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000327,328/2014003-01 NCV Failure to Perform Visual Examination of the

Unit 1 and Unit 2 CRDM Seismic Plate

Supports (Section 1R08)

05000327,328/2014003-02 NCV Failure to Comply with Entry requirements

to a HRA. (Section 2RS1)

Closed

2515/182 TI Review of the Industry Initiative to Control

Degradation of Underground Piping and

Tanks, Phase II (Section 4OA5.1)

Attachment

LIST OF DOCUMENTS REVIEWED

Section R04: Equipment Alignment

Procedures

0-GO-16, System Operability Checks, Rev 19

Section R05: Fire Protection

Procedures

SQN-FPR-Part-II, SQN Fire Protection Report Part II - Fire Protection Plan, Revision 28

Other documents

TUR-0-706-01, Fire Protection Pre-Fire Plans Turbine Building - El. 706, Revision 3

TUR-0-706-02, Fire Protection Pre-Fire Plans Turbine Building - El. 706, Revision 3

CON-0-706-00, Fire Protection Pre-Fire Plans Control Building - El. 706, Revision 6

CON-0-706-00, Fire Protection Pre-Fire Plans Control Building - El. 732, Revision 7

Section R06: Flood Protection Measures

Other documents

TVA letter to NRC dated May 4, 2007. TVA response to GL 2007-01

48N1225

47W852-1

SQN-SQS4-0056, Moderate Energy Line Break Flooding Study, Revision 2

Section R08: Inservice Inspection

Drawings

CCD No. 2-2H63-0006-01, Residual Heat Removal System Pipe Support, Rev. 0

CCD No. 2-2-H2O-0020-01, Sequoyah Nuclear Plant, Safety Injection System Pipe Supports,

R-0

1095E46, Sequoyah Nuclear Plant No. 2, CRDM Seismic Support Platform General Assembly,

Sheet 1 of 2, Rev. 6

1, 2-47W813-1-ISI, Flow Diagram Reactor Coolant System, Rev. 7

2-47K406-112, N2-62-12A Isometric, Static, Thermal, and Seismic Analysis of CVCS Piping,

Rev.1

ISI-0401-C-02, Sequoyah Nuclear Plant Unit-2 Steam Generator Replacement, Rev. 3

48N427, Structural Steel Equipment Supports Upper Steam Generator Support, Rev. 15

48N431, Structural Steel Equipment Supports Upper Steam Generator Support Details, Rev. 20

DCA No. D22479-3001, Drawing Change Authorization, DCN D22479A, Page 85, Rev. 2

DCA No. D22479-3002, Drawing Change Authorization, DCN D22479A, Page 86, Rev. 1

DCA No. D22479-3003, Drawing Change Authorization, DCN D22479A, Page 87, Rev. 2

DCA No. D22479-3005, Drawing Change Authorization, DCN D22479A, Page 89, Rev. 2

CCD No. 2-2-H47-0104-01, Steam Generator Blowdown System Pipe Supports, Rev. 0

Procedures

0-PI-SLT-068-200.0, Reactor Building Post Shutdown Leakage Examination, Rev. 4

2-PVC-068-0340B, Preventative Maintenance Work Instruction, PM011442000, Attachment A

used by WO 114734912, dated 5/12/14

0-MI-MRR-068-006.0, Sequoyah Nuclear Plant, Installation of Reactor Pressure Vessel Head

and Attachments, Appendix G Seismic Tie Data Sheet, Rev. 38

Attachment

3

2-SI-SXI-068-201.0, Leakage Test of the Reactor Coolant Pressure Boundary, 1/21/2014

N-PT-9, Liquid Penetrant of ASME and ANSI Code Components and Welds, Rev. 0037

N-VT-1, Visual Examination Procedure for ASME Section XI Preservice and Inservice, Rev. 5

Areva 03-9052292, Operating Instructions for RANGER in recirculating Steam Generator,

Rev. 15

Areva 03-9187284, Utilizing a Personal Computer Platform for Eddy Current Acquisition Data

Functions, Rev. 1

NEDP-16, Steam Generator Program, Rev. 15

0-MI-MXX-068-005.0, Steam Generator Primary Side Maintenance Activities, Rev. 24

0-MI-MXX-003-002.0, Steam Generator Secondary Side Maintenance Activities, Rev. 15

0-SI-SXI-068-114.3, Steam Generator Tubing Inservice Inspection and Augmented Inspections,

Rev. 1

2-SI-CEM-068-137.5, Primary to Secondary Leakage via Steam Generators, Unit 2, Rev. 13

Engineering/Technical Evaluations

PER 888330, Boric Acid Leakage Evaluation, Reactor Cavity Nozzle Cover Seal leaking into

Keyway, 6/24/14

PER 890230, Evaluation of Boric Acid Corrosion Damage, 2-SNUB-068-RCH30, 6/7/14

Sequoyah PER 743110, Degraded Unit 2, Ice Condenser Due To Recurring Frost Accumulation

on Intermediate Deck Doors, Event Date, 5/08/13

SQN PER 889645, Equipment Apparent Cause Evaluation for Compression Fitting Leak, Event

Date, 6/24/14

NOI-2-SQ-432, Available Margins in Pipe Support Attributes, 5/26/14

Corrective Action Documents

PER 888991, Observation made during NRC ISI - Boric Acid Inspection, 5/28/14

PER 889400, Determine whether CRDM Seismic Support should be examined under Section

XI, 5/23/14

PER 899941, Failure to Quarantine Failed Part for Analysis, 6/17/2014

PER 743110, Degraded Non-conforming condition for PDO on RCS leakage and Ice on

Intermediate deck doors (IDD), 7/19/13

SR888431, Loose Hydraulic lines on Snubbers, 5/22/2014

PER 487507, SQN review/Westinghouse NSAL-12-1 SG Channel Head Degradation, 2012

PER 889451, Discoloration in Steam Generator Primary Bowls, 2014

SR 890656, Steam Generator Secondary Side Inspection and Sludge Lancing

SR 891631, EPRI ETSS not referred in site ETSS, 2014

SR 891633, Steam Generator ECT Secondary Analyst did not call wear out and proximity

indications, 2014SR 900540, Evaluate SEQ Primary to Secondary Leakrate Detection Limits,

2014

Other Documents

Penetration Number 56, RPV Head Penetration UT Data Sheet, 12/7/06

Penetration Number 56, RPV Head Penetration UT Data Sheet, 05/18/14

Penetration Number 60, RPV Head Penetration UT Data Sheet, 05/18/14

Penetration Number 53, RPV Head Penetration UT Data Sheet, 12/06/06

Penetration Number 53, RPV Head Penetration UT Data Sheet, 05/19/14

R-6069, TVA Record of Liquid Penetrant Examination, 2SIH-020-IA, 4/30/99

Attachment

4

R0114, TVA Liquid Penetrant Examination, Reinspection Summary No. 01961-ISI-SQN,

for 2-SIH-020-IA, 5/19/14

R0105, TVA Liquid Penetrant Examination, Inspection Summary No. 01934-ISI-SQN2,

or 2-CVCH-006-IA, 5/18/14

NPG-SPP-09.1, ASME Code and Augmented Programs, Attachment 8, Form NPG-SPP-09.1-2,

for Component ID, 2-SIH-020-IA, 5/19/14

NPG-SPP-09.1, ASME Code and Augmented Programs, Attachment 8, Form NPG-SPP-09.1-2,

for Component ID, 2-CVCH-585, 5/12/14

System 068, Reactor Coolant System Health Report, 2/1/2014 - 5/31/2014

0-SI-DXI-000-114.3, Attachment 5, Unit-2 Examination Schedule for ASME Class 1, 2, 3

Components, 5/9/14

TVA Report No. R0105, Summary No. 01934-ISI-SQN2, Liquid Penetrant Examination

Summary for Component ID 2-CVCH-006-IA, Category B-K/B10.20, Integral Attachment,

5/18/14

Work Order No. 112354373, Valve SQN-2-VLV-001-0817 Replacement, 3/25/13

R0041, TVA Record of Visual Examination, 2-CVCH-585, 5/6/2014

R0086, TVA Record of Visual Examination, 2-CVCH-584, 5/15/2014

R0094, TVA Record of Visual Examination, 2-CVCH-586, 5/14/2014

R0151, Ultrasonic Piping Examination Data Sheet, FDF-011A, 5/24/2014

R0152, Ultrasonic Piping Examination Data Sheet, FDF-010C, 5/24/2014

R0170, TVA Record of Visual Examination, SGH-4-1, 5/28/14

Candidate No. 3237861, EPRI Performance Demonstration Initiative Program Qualifications,

1/14/11

MWK7861, IHI Southwest Technologies Inc. Certificate of Qualification, 2/22/2013

H14132981, Certificate of Calibration, M&TE ID No. E41820, 4/14/2013

VT-1, Certificate of Method Qualification Record for BMNO6QGPV, Expires, 11/28/2014

VT-3, Certificate of Method Qualification Record for BMNO6QGPV, Expires, 11/28/2014

VT-3, Certificate of Method Qualification Record for D880WSO0D, Expires, 10/5/2014

0-SI-MFT-000-001.0, Appendix E Page 1, Snubber Functional Testing, SQN-2-SNUB-015-

SGBH104, 5/18/14

Report No. SCV-0001, Visual Examination of IWE Interfaces, Moisture Barrier, 5/20/2014

Report No. SCV-0004, Visual Examination of IWE Interfaces, Moisture Barrier, 5/5/2014

Report No. SCV-0005, Visual Examination of IWE Interfaces, Moisture Barrier, 5/5/2014

0-TI-DXX-000-097.1, Boric Acid Corrosion Control Program, Rev. 0009

NPG-SPP-09.7.4, Boric Acid Corrosion Control Program, Rev. 0001

0-TI-SPT-000-301.0, ASME Section XI Pressure testing Program Basis Document, Rev. 0004

0-TI-RVI-000-301.0, Sequoyah Unit 1 & 2, PWR Reactor Vessel Internals Inspection Program,

Rev. 0

Sequoyah Unit 2 Control Room, Total Unidentified Leakage Logs, 12/28/2012 thru 5/10/2014

Areva Use of Appendix H and Appendix I Qualified Techniques Sequoyah U2R19 Refueling

Outage, Rev. 0, May 2014

Areva 51-91988290-00, Sequoyah U2R19 Steam Generator Degradation Assessment,

Rev. 0, May 2014

Areva SQN 2C19 Analyst Training Instructions, Rev. 0

Areva 54-ISI-400-021, Eddy Current Inspection Multi-Frequency Eddy Current Examination of

Tubing, June 2013

Areva 51-9221442-000, Sequoyah Unit 2 EOC19 SG ECT Inspection Plan

Areva ETSS_BOB1, Areva Examination Technical Specification Sheet for Bobbin Probe, Rev. 0

Attachment

5

Areva ETSS_RPC1, Areva Examination Technical Specification Sheet for MRPC Probe, Rev. 0

Areva ETSS_RPC2, Areva Examination Technical Specification Sheet for MRPC Probe, Rev. 0

Areva ETSS_Array1, Areva Examination Technical Specification Sheet for MRPC Probe, Rev. 0

B85 130429 005, Sequoyah Unit 2 Refueling 18 Operational Assessment, Rev. 1

Calibration Records for Eddy Current Tester Miz80i Serial Numbers: 39, 21, 71, 36, 73, 91

Certificate of Conformance for Eddy Current Probes, Serial Numbers 653790, 652262, 655350,

653784, 652251, 653863, 652265, 652242

Calibration Standard for ASME 21095, 21099, 21100, 21096, EDM 9173936, 21086,

ARRAY 9173939

Personnel Qualification Records for Qualified Data Analysts: W. Bridforth, D. Cornell,

N. Farenbaugh, J. Janet Sr, R. Lee, G. Manley, W. McMillan, S. Merriam, E. Miranda,

R. Miranda, J. Parrish, J. Oliver, A. Richardson, T. Shulter, J. Sordini, L. Tobin, D. Torres

Personnel Qualification Records of TVA Steam Generator Program Personnel: J. Mayo,

W. James

SQN-ENG-F-10-02, Self-Assessment on Steam Generator Program, April 2010

SQN-ENG-S-11-91, Benchmarking Report on U2R17 NRC Inservice Inspection Readiness,

March 2011

SQN-CEM-S-10-015, Self-Assessment on EPRI Secondary Water Chemistry Guidelines,

July 2010

Sequoyah Nuclear Plant Unit 2, Replacement Steam Generator Eddy Current Examination

Guideline, Rev. 1

Structural Integrity Associates, Report No. 1400660.401.R0, Independent Review of

Westinghouse LTR-SGMMP-14-27, Assessment of Discolorations on Replacement Steam

Generator Channel Head Cladding at Sequoyah Unit 2, dated May 30, 2014

Section R12: Maintenance Effectiveness

Procedures

TI-4, Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting -

10CFR50.65, Revision 25

Section R13: Maintenance Risk Assessments and Emergent Work Evaluation

Procedures

NPG-SPP-07.0, Work Management, Revision 0

NPG-SPP-07.1, On Line Work Management, Revision 9

NPG-SPP-07.3, Work Activity Risk Management Process, Revision 13

NPG-SPP-07.11.1, Equipment Out of Service Management, Revision 9

Section R15: Operability Evaluations

Procedures

NEDP-22, Functional Evaluations, Rev. 15

OPDP-8, Limiting Conditions for Operation Tracking, Rev. 16

NPG-SPP-03.5, Regulatory Reporting Requirements, Rev. 10

Section R19: Post Maintenance Testing

Procedures

MMDP-1, Maintenance Management System, Rev. 20

MMDP-3, Guidelines for Planning and Execution of Troubleshooting Activities, Rev. 6

NPG-SPP-6.5, Foreign Material Control, Rev. 4

Attachment

6

NPG-SPP-6.1, Work Order Process Initiation, Rev. 2

NPG-SPP-06.3, Pre-/Post-Maintenance Testing, Rev. 1

NPG-SPP-06.9, Testing Programs, Rev. 0

NPG-SPP-06.9.1, Conduct of Testing, Rev. 8

NPG-SPP-06.9.3, Post-Modification Testing, Rev. 5

Section 2RS1: Radiological Hazard Assessment and Exposure Controls

Procedures, Guidance Documents, and Manuals

NPG-SPP-05.1 Radiological Controls Revision (Rev.) 003

NPG-SPP-05.1.1, Alpha Radiation Monitoring Program, Rev. 003

NPG-SPP-05.6, Controlling Byproduct and Source Material, Rev. 002

O-SI-RCI-000-056.0, Byproduct Material Inventory and Sealed Source Leak Test Rev. 016

RCI-14, Radiation Work Permit (RWP) Program, Rev. 058

RCI-15, Radiological Postings Rev. 026

RCI-21, Control of Radioactive Materials, Rev. 019

RCI-22, Contamination Control Rev. 024

RCI-24, Control of Very High Radiation Areas Rev. 014

RCI-28, Control of Locked High Radiation Areas Rev. 015

RCI-29, Control of Radiation Protection Keys, Rev. 016

RCI-201, Radiation and Contamination Surveys, Rev. 015

RCI-202, Airborne Radioactivity Surveys Rev. 008

RCI-204, Radiological Surveys of Equipment and Materials Leaving the RCA, Rev. 008

RCI-404, Radiation Protection Requirements for Remote Job Coverage, Rev. 001

RCI-417, Radiological Monitoring of the Hydrogen Peroxide Injection Crud Burst, Rev. 001

RCDP-1, Conduct or Radiological Control Rev. 005

Records and Data

0-SE-RCI-000-056.0 Byproduct Material Inventory and Sealed Source Leak Test, 11/23/2013

0-SE-RCI-000-056.0 Byproduct Material Inventory and Sealed Source Leak Test, 05/02/2014

0-TI-NUC-000-002.0, Storing Material in Spent Fuel Pool or New Fuel Vault, Rev. 0022,

Attachment-1, Inventory of Materials in Spent Fuel Pool, Cask Loading Area, and New Fuel

Vault, dated 02/06/2014.

2013 Sequoyah Radiation Protection Annual Report conducted per NPG-SPP-05.1,

Radiological Controls, Section 3.2, Program Monitoring Evaluation and Oversight

Airborne Radiation Survey (ARS) Number (No.) 051514008, U-2 Lower Seal Table During Eddy

Current Test, dated 05/15/2014

ARS No. 051514004, U-2 Lower Containment Raceway Routine, dated 05/15/2014

ARS No. 051514002, U-2 Lower Containment Routine @Elevation 679, dated 05/15/2014

ARS No. 051514003, U-2 Lower Containment Routine @IPCW, dated 05/15/2014

ARS No. 052114017, U-2 S/G#1 Laydown Area During Insert Removal, dated 05/21/2014

ARS No. 052114018, U-2 S/G#1 During Insert Removal, dated 05/21/2014

ARS No. 052114019, U-2 S/G#4 During Insert Removal, dated 05/21/2014

ARS No. 052114022, U-2 S/G#3 During Insert Removal, dated 05/21/2014

ARS No. 052114023, U-2 S/G#2 During Insert Removal, dated 05/21/2014

ARS No. 052114024, U-2 S/G#3 Laydown Area During Insert Removal, dated 05/21/2014

ARS No. 052114025, U-2 S/G#1 Laydown Area Back-up Sample, dated 05/21/2014

ARS No. 052114026, U-2 S/G#3 Laydown Area Back-up Sample, dated 05/21/2014

ARS No. 052114027, U-2 S/G#3 Primary PlatformBack-up Sample, dated 05/21/2014

Attachment

7

Annual Inventory Reconciliation Confirmation for License #DRP-77, dated 01/14/2014

NPG Daily Outage Report, dated 05/12/2014

NPG Daily Outage Report, dated 05/13/2014

RWP No. 14000063, LHRA - Plant Filter Change Outs: Seal Water Injection and Return, RCS,

SFP, SFP Skimmer, Ion Exchange Filters and Refuel Water Purification Filters: Change Out and

Transport.

RWP No. 14220052, U2 Lower Containment: IPCW - (HRAs) - MOVATs Testing

RWP No. 14220103, U2 Lower Containment, Excess LTDWN. HX. RM - LHRA

RWP No. 14220122, U2 Lower Containment - Seal Table Work to Include Table Roll Back,

Tube Extraction, High Pressure Seals, Install Ferrules, Tube Cutting and Recovery Efforts

RWP No. 14230013, U2 Lower CTMT - Steam Generator Primary Platforms -LHRA

RWP No. 14230023, U2 Lower Containment - Steam Generator - LHRA - Full Jump for

Installing/Removing Nozzle Dams

Survey No. SQN-M-20140516-20, U2 Upper Containment - All Areas, 05/16/2014

Survey No. SQN-M-20140527-16, Reactor head move from cavity to head stand, 05/27/2014

Survey No. SQN-M-20140514-1, U-2 Raceway Elev 679, 05/14/2014

Survey No. SQN-M-20140515-4, U-2R19 Lower IPCW Floor Area, 05/15/2014

Survey No. SQN-M-20140515-7, U2 #2 RCP Platform, 05/15/2014

Survey No. SQN-M-20140521-22, U2 Steam Generator (S/G) Platform, 05/21/2014

Survey No. SQN-M-20140522-6, U2R19 S/G 1&4 S/G Generator Platform, 05/22/2014

Survey No. SQN-M-20140605-4, 5&6, SQN ISFSI PAD Quarterly Routine, 06/05/2014

Survey No. SQN-M-2014021-3, 5&6, SQN ISFSI PAD Quarterly Routine, 02/16/2014

Survey No. SQN-M-20131020-15, SQN ISFSI PAD Quarterly Routine, 10/20/2013

Corrective Action Program (CAP) Documents

PER-661017

PER-713213

PER-776043

PER-776044

PER-790597

PER 805944

PER-805952

PER 807919

PER 827948

PER 868727

PER 881321

PER 886160

PER 886668

PER 888770

Section 2RS2: Occupational ALARA Planning and Controls

Procedures, Guidance Documents, and Manuals

CHEM -002, Primary Water Chemistry Program Strategic Plan, Rev. 6

NPG-SPP-05.2, ALARA Program, Rev. 4

NPG-SPP-05.2.1, Operational ALARA Planning and Controls, Rev. 2

NPG-SPP-05.2.2, Establishing Collective Radiation Exposure Annual Business Plan Goals,

Rev. 0

NPG-SPP-05.2.3, Outage Exposure Estimating and Tracking, Rev. 0

Attachment

8

RCI-10, ALARA Program, Rev. 35

RCI-14, Radiation Work Permit (RWP) Program, Rev. 58

RCI-19, Temporary Shielding Program, Rev. 13

RCI-417, Radiological Monitoring of the Hydrogen Peroxide Injection Crud Burst, Rev. 1

Reports, Records, and Data

ALARA Committee Meeting Minutes - Meeting Number (No.) 2013-04, 2/22/2013

ALARA Committee Meeting Minutes - Meeting No. 2013-11, 7/11/2013

ALARA Committee Meeting Minutes - Meeting No. 2013-19, 10/17/ 2013

ALARA Committee Meeting Minutes - Meeting No. 2013-22, 11/14/ 2013

ALARA Plan: 2013-010, Refueling Operations

ALARA Plan: 2013-011, Mechanical Maintenance Group (MMG)

ALARA Plan: 2013-012, Electrical Maintenance and RCPs

ALARA Plan: 2013-015, Plant Services

ALARA Plan: 2013-017, Radiation Protection

ALARA Plan: 2013-018, U1R19 MODS Ice Condenser/Snubbers/Insulation/Scaffolds/Painting

ALARA Plan: 2014-010, Refueling Operations

ALARA Plan: 2014-011, Mechanical Maintenance Group (MMG)

ALARA Plan: 2014-015, Plant Services (RCL)

ALARA Plan: 2014-017, Radiation Protection

ALARA Plan: 2013-018, Modifications U2R19

ALARA Work in Progress Review: RWP 2013-011, 10/24/13

ALARA Work in Progress Review: RWP 2013-018, 11/14/13

Fiscal Year (FY) Dose Estimate Approval per NPG-SPP-5.2.2 3.1.2

FY14 RP Dose Reduction Plan

Graphic Showing Co-58 and Co-60 Trends for 2013-2024 in Both Units

Report Sequoyah TEDE Year to Date as of 4/24/14

SQN U1R19 Outage - Dose Reduction Plan

Slide Show: Sequoyah Nuclear Plant 2011-2015 Business Plan: Collective Radiation Exposure

2014-2018 SQN Business Plan

U1R19 ALARA OUTAGE REPORT

U1R19 Final ALARA Plan Status

U2R19 ALARA Plan Challenge Numbers Spreadsheet

CAP Documents

2013 Sequoyah Radiation Protection Annual Report, 10/25/2013

Audit SSA1309, Radiation Protection Sequoyah Nuclear Plant, 8/19-30/2013

PER 773873

PER 776064

PER 776639

PER 770709

PER 773258

PER 724010

PER 798963

PER 801067

PER 886820

PER 853897

Attachment

9

Section 2RS3: In-Plant Airborne Radioactivity Control And Mitigation

Procedures and Guidance Documents

0-PI-FPU-049-401.M, Self Contained Breathing Apparatus, Rev. 030

0-PI-RCI-033-001.0, Periodic Monitoring of Service Air System for Use as Breathing Air,

Rev. 008

NPG-SPP-05.10, Radiological Respiratory Protection Program, Rev. 003

RCI-04.01, Selection, Issue, and Use of Respiratory Protection Devices, Rev. 008

RCI-04.02, Cleaning/Sanitizing, Maintenance, Inspection, Storage and Inventory or Respiratory

Protection Devices, Rev 4

RCI-04.03, Respiratory Protection Program Periodic Evaluation Rev. 000

RCI-18.01, DOP Testing of Portable HEPA and Vacuum Cleaners, Rev 001

UFSAR Chapter 11 & 12

Records and Data Reviewed

AIR/GAS Quality Report and Certificates for SN: 11040, Kit #279317, dated 10/01/2013 and Kit

  1. 286878, dated 03/20/2013

Assessment SQN-RP-S-14-003, In-Plant Airborne Radioactivity Control and Mitigation,

01/15/2014

Grade D Certificates for Plant System Air Compressor Equipment ID#s 0-CLR-32-25, 0-CLR-

32-26, 0-CLR-32-27 and 0-DS-32-136, dated 09/25/2013

Grade D Certificates for Plant System Air Compressor Equipment ID#s 0-CLR-32-25, 0-CLR-

32-26, 0-CLR-32-27 and 0-DS-32-136, dated 02/13/2012

HEPA DOP Test Certification for Vacuum Cleaner #s TVA-2 and 1369, dated 05/14/2014

KeyStone Certifications for Vacuum HEPA Filters #2801, #2787 and #2790, dated 03/13/2014

MSA MMR Certification Records for TVA SCBA Repair Technicians, Current

MSA Posi3 USB Complete SCBA Test Results for Units CR06, CR12 and CR13, dated

08/06/2013

Personnel Contamination Log, 1/2013-5/2014

UNITECH Services Group DOP Test Results for HEPA #700-7, dated 12/09/2014

UNITECH Services Group DOP Test Results for HEPA #700-8, dated 11/01/2013

UNITECH Services Group DOP Test Results for HEPA #700-29, dated 10/02/2013

CAP Documents

PER 660950

PER 805989

Section 2RS4: Occupational Dose Assessment

Procedures and Guidance Documents

NPG-SPP-05.1.1 Alpha Radiation Monitoring Program, Rev. 003

RCI-05.304, WBC Routine Operations and DAC-Hr Assignment Evaluation, Rev. 009

RCI-202, Airborne Radioactivity Surveys, Rev. 008

RCI-209, Radiological Surveys of Personnel Leaving the RCA or Protected Area, Rev. 004

RCDP-7, Bioassay and Internal Dose Program, Rev. 005

RCDP-10, Personnel Contamination Reporting, Rev. 005

RCTP-106 Special Dosimetry Operations, Rev. 003

RCTP-113, External Dosimetry MQA Program, Rev. 000

Attachment

10

Records and Data Reviewed

Assessment SQN-RP-S-14-004, Occupational Dose Assessment, 02/10/2014

Committed Effective Dose Equivalent Assignment Summary for 2013

Dosimetry Investigation Reports 2014-015, 2014-016 and 2014-017

Evaluation of the Canberra GEM-5 Portal Contamination Monitor Detection Capabilities for Use

as a Passive Whole Body Count Instrument, dated 10/30/2014

Investigative Whole body Counts (6) for Intakes Occurring on U2 S/G Platform on 05/21/2014

Multi-Badge EDEX Worksheet for Entry on RWP No. 14240182, dated 05/15/2014

Multi-Badge EDEX Worksheet for Entry on RWP No. 14240213, dated 05/14/2014

Multi-Badge EDEX Worksheet for Entry on RWP No. 14240053, dated 05/14/2014

Multi-Badge EDEX Worksheets (2) for Entries on RWP No. 14240023, dated 05/23/2014 and

05/29/2014

NVLAP Certification of Accreditation to ISO/IEC 17025-2005 for 2014

OSL Dosimetry Investigation Summary for 01/2013-05/2014

SQN TLD Area Monitoring Results for 4th Qtr. 2013

CAP Documents

PER 675250

PER 753263

PER 784430

PER 798104

PER 829995

PER 830008

PER 845120

PER 857054

PER 869683

PER 881323

PER 888629

PER 888987

Section 2RS5: Radiation Monitoring Instrumentation

Procedures and Guidance Documents

1-SI-ICC-090-400.0, Calibration of Shield Building Vent Radiation Monitor 1-RM-90-400,

Rev. 18

2-SI-ICC-090-400.0, Calibration of Shield Building Vent Radiation Monitor 2-R-90-400, Rev. 18

CHTP-109, Chemistry QA/QC, Rev. 8

EPIP-1, Emergency Plan Classification Matrix, Rev. 50

NPG-SPP-06.7, Instrumentation Setpoint, Scaling and Calibration Program, Rev. 2

RCI-5, Radiation Protection Instrumentation Program, Rev.77

RCI-5.100, Operation of Laboratory Counter/ Scalers, Rev. 6

RCI-5.102, Calibration and Operation of the Canberra iSOLO Model 300G Alpha/Beta Counter,

Rev. 5

RCI-5.300, Calibration and Operation of the Eberline Personnel Contamination Monitor (PCM-

1B), Rev. 3

RCI-05-301, Operational Checks for the GEM-5 Portal Monitor, Rev. 8

RCI-05.305, Calibration, Response Check, And Operation of the Canberra ARGOS-5AB

Personnel Contamination Monitor, Rev.7

Attachment

11

RCI-05.306, Calibration, Response Check, and Operation of the Canberra Cronos-4and

Cronos-11 Contamination Monitors, Rev.3

RCI-05.400, Criteria for Setting Portable Radiation Protection Instrument Response Check

Windows, Rev.4

RCI-05.408, Response Check of Neutron Survey Instruments, Rev. 0

RCI-05.401, Instrument Response Checks Utilizing the Shepherd Calibrator, Rev.4

Records and Data Reviewed

Apex Gamma Spectroscopy Efficiency Calibration, Detector 2, 1/24/2012

Calibration Data Records for the following instruments:

ARGOS-5AB, TVA# 860588, 5/8/2013 and 4/30/2014

ARGOS-5AB, TVA# 860589, 4/26/2013 and 4/9/2014

Bicron Analyst [no probe type specified], TVA# 8355305, 7/30/2013

Bicron Analyst with GM, TVA# 8355305, 2/4/2014

Bicron Analyst with NaI, TVA# 835539, 6/4/2103 and 1/7/2014

Cronos, TVA# 860780, S/N 1203-021, 2/5/2013 and 1/27/2014

Eberline Teletector, TVA# 523331, 8/28/2013 and 2/26/2014

Eberline Teletector, TVA# 523338, 8/22/2013 and 2/26/2014

GEM-5, S/N 1203-021, 5/7/2013 and 4/23/2014

GEM-5, S/N 0909-179, 3/18/2013 and 3/14/2014

HV-1 [air sampler], TVA# 556318, 10/1/2013 and 3/24/2014

HV-1 [air sampler], TVA# 860003, 8/30/2013 and 3/27/2014

Ludlum Model 3 frisker, TVA# 860888, 5/30/2013 and 2/18/2014

Ludlum 2200, TVA# 860654, 10/17/2012 and 12/3/2013

Ludlum 3030P, TVA# 951047, 12/17/2014

Ludlum 9-3, TVA# 860844, 5/22/2013 and 2/4/2014

Ludlum 9-3, TVA# 861000, 2/25/2013 and 2/5/2014

MG Telepole WR, TVA# 860096, 3/5/2013 and 1/7/2014

MG Telepole WR, TVA# 951056, 3/3/2014 and 5/14/2014

SAIC H-810, TVA# 838786 and TVA# 860029, 6/5/2013 and 2/24/2014

SAM-11, TVA# 860323, 6/17/2013 and 3/10/2014

SAM-11, TVA# 860324, 2/4/2013 and 1/30/2014

PCM-1B, TVA# 576358, 8/16/2012 and 7/26/2013

PCM-1B, TVA# 484689, 4/19/2012 and 9/23/2013

Calibration/Efficiency Check, Tri-Carb Model 3100TR, S/N 060450, 6/2/2014

Calibration Report, Calibration of the FASTSCAN 1 WBC System at the Dosimetry Lab of the

TVA Sequoyah Nuclear Plant, 8/22/2013

Calibration Report, Calibration of the FASTSCAN 2 WBC System at the Dosimetry Lab of the

TVA Sequoyah Nuclear Plant, 8/22/2013

Calibration Report, High Range Radiation Monitor Calibrator RT-11, S/N 24, TVA Source #

775N, 3/15/1983

Certificate of Calibration, Beta Standard Source, S/N G4-973, TVA Source # 2482, 11/15/2009

Certificate of Calibration, Beta Standard Source, S/N G4-972, TVA Source # 2485, 11/15/2009

Certificate of Calibration, Standard Radionuclide Source, S/N 86078-166, 10/1/2011

Certificate of Calibration, Standard Radionuclide Source, S/N 95460, 1/1/2014

Certificate of Gamma Standard Source, S/N 205-56-4, TVA Source # 1295N, 8/1/1989

Certificate of Gamma Standard Source, S/N 363-02-3, TVA Source # 1296N, 5/1/1990

Certificate of Gamma Standard Source, S/N 205-83-5, TVA Source # 1297N, 1/1/1990

Attachment

12

Certificate of Gamma Standard Source, S/N M-246, TVA Source # 1297N, 5/15/1990

Certificate of Gamma Standard Source, S/N M-250, TVA Source # 1299, 5/15/1990

Certificate of Gamma Standard Source, S/N M-248, TVA Source # 1300N, 5/15/1990

Certificate of Gamma Standard Source, S/N 349-29-1, TVA Source # 1301, 8/1/1989

Digital Air Flow Calibrator, TVA# 860169, S/N 3204, Source Check Record, 11/23/13

F & J Specialty Products, Inc. Certificate of Calibration, Digital Calibrator Model D-828B, Serial

  1. (S/N) 3204, 11/5/2013

Sequoyah Offsite Dose Calculation Manual (ODCM), Rev. 58

Source Response and Background Data Sheet, Ludlum 2200 Scaler, TVA# 860654, February

and March 2014

System Health Report, System 90, Radiation Monitoring, 10/1/2013 through 1/31/2014

White Paper, Waste Stream Analysis (DAW 10/14/2013), 3/15/2014

White Paper, Sequoyah Whole Body Counter Library Revision, 5/30/2014

Whole Body Counter Library Listing, Europium -152, 6/4/2014

Work Order (WO) No. 112807041, 1-SI-ICC-090-400.0 Shield Building Vent Rad Mon 1-RM-90-

400 Cal, 10/10/2012

WO No. 115052959, 1-PI-CEM-043-487.0 U1 Post Accident Sampling Sys Calibration,

1/29/2014

WO No. 112625727, 0-SI-ICC-090-101.B Aux Bldg Vent Gaseous Rad Mon 0-R-90-101B &

Flow Monitor 0-F-30-174 CC, 8/31/2012

WO No. 114475841, 0-SI-ICC-090-101.B Aux Bldg Vent Gaseous Rad Mon 0-R-90-101B &

Flow Monitor 0-F-30-174 CC, 9/6/2013

CAP Documents

Assessment SQN-RP-S-14-002, RP Portable Instrumentation and Calibration, 1/28 to 1/30/2014

PER 735601

PER793878

PER 801879

PER 832856

PER 871912

Section 4OA1: Performance Indicator Verification

Procedures

NPG-SPP-02.2, Performance Indicator Program, Rev. 2

NPG-SPP-02.2, Performance Indicator Program, Rev. 6

NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Rev. 6

Records and Data Reviewed

2013 Annual Radioactive Effluent Release Report, 4/16/2014

2013 Annual Radiological Environmental Operating Report, 4/16/2014

Gaseous Radioactive Waste Release Permit 2014027.059.014.G

Liquid Radioactive Waste Release Permit 2013191.007.087.L

CAP Documents

PER 756809

PER 824084

PER 788604

PER 793921

Attachment

13

Section 4OA5: Temporary Instruction 2515/182 - Review of the Industry Initiative to

Control Degradation of Underground Piping and Tanks

Corrective Action Program Documents

PER 175149-002, 4 Diesel Fuel Oil Line Failed Pressure Test

PER 347970, NEI 09-14, NSIAC Buried Piping Completion Dates To Be Assigned as LTCAs

PER 684460, License Renewal NRC Commitment #3: Revise the Buried and Underground

Piping and Tanks Inspection Program to Meet License Renewal Requirements

Procedures

0-PI-DXX-000-750, Piping Inspection in Tunnels and Infrequently Accessed Areas, Rev. 0000

0-SI-SXI-067-300.7, System Leakage Test of the Essential Raw Cooling Water System Buried

Piping, Rev. 0002

0-TI-DXX-000-915.0, Underground Piping and Tanks Integrity Program, Rev. 0006

G-55, Technical and Programmatic Requirements for the Protective Coating Program for TVA

Nuclear Plants, Rev. 19

G-94, Piping Installation, Modification and Maintenance, Rev. 2

NPG-SPP-09.15, Underground Piping and Tanks Integrity Program (UPTI), Rev. 0006

Other Documents

0901186.000, Structural Integrity Associates, Inc. Baseline Risk Implementation Analysis:

Sequoyah Nuclear Power Plant, Rev. 0

1200931.401, Sequoyah Nuclear Plant Buried Piping Cathodic Protection Design Study, Rev. 0

Buried Pipe Integrity Program Corrosion Assessment for Buried Piping Systems, dated

February 2010

CRP-ENG-F-12-002, Assessment of the Underground Piping and Tanks Integrity Program

SQN-ENG-S-14-016, Self-Assessment: Readiness for NRC TI 2515/182 Phase 2 Inspection

Underground Piping and Tanks Integrity Program Inspection Plan, Rev. 3, dated April 1, 2014

WO 09-777416-005, Perform UT Examination for Wall Thickness of Excavated Bare Metal on

Diesel Fuel Oil Line

Attachment

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