ML14213A166
| ML14213A166 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| 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) | |
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
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
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
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
- 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
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.
- 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.
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
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.
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
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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
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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
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.
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- 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
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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
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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:
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
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
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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
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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
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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
- 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
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 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
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
- 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
- (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