ML14038A346
| ML14038A346 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 02/07/2014 |
| From: | Bartley J H Reactor Projects Region 2 Branch 6 |
| To: | Shea J W Tennessee Valley Authority |
| References | |
| IR-13-005 | |
| Download: ML14038A346 (44) | |
See also: IR 05000327/2013005
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II 245 PEACHTREE CENTER AVENUE NE, SUITE 1200 ATLANTA, GEORGIA 30303-1257
February 7, 2014
Mr. Joseph W. Shea
Vice President, Nuclear Licensing
Tennessee Valley Authority
Chattanooga, TN 37402-2801
SUBJECT: SEQUOYAH NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT
05000327/2013005 AND 05000328/2013005
Dear Mr. Shea:
On December 31, 2013, the U.S. Nuclear Regulatory Commission (NRC) completed an
inspection at your Sequoyah Nuclear Plant, Units 1 and 2. On January 13, 2014, the NRC
inspectors discussed the results of this inspection with Mr. Carlin and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report.
NRC inspectors documented one self-revealing finding of very low safety significance (Green) in this report. This finding involved a violation of NRC requirements. The NRC is treating this
violation as a non-cited violation (NCV) consistent with Section 2.3.2.a of the NRC Enforcement
Policy. If you contest the violation or significance of this 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, D.C. 20555-0001; with
copies to the Regional Administrator, Region II; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 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.
As a result of the Safety Culture Common Language Initiative, the terminology and coding of
cross-cutting aspects were revised beginning in calendar year (CY) 2014. New cross-cutting
aspects identified in CY 2014 will be coded under the latest revision to Inspection Manual Chapter (IMC) 0310. Cross-cutting aspects identified in the last six months of 2013 using the
previous terminology will be converted to the latest revision in accordance with the cross-reference in IMC 0310. The revised cross-cutting aspects will be evaluated for cross-cutting
themes and potential substantive cross-cutting issues in accordance with IMC 0305 starting with
the CY 2014 mid-cycle assessment review.
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
NRC's Public Document Room or from the Publicly Available Records (PARS) component of NRC's Agencywide Documents Access and Managem
ent System (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Jonathan H. Bartley, Chief Reactor Projects Branch 6
Division of Reactor Projects
Docket Nos.: 50-327, 50-328 License Nos.: DPR-77, DPR-79
Enclosure: Inspection Report 05000327/2013005, 05000328/2013005 w/Attachment: Supplementary Information
cc: via ListServ distribution
_________________________ SUNSI REVIEW COMPLETE FORM 665 ATTACHED OFFICE RII:DRP RII:DRP RII:DRS RII:DRS RII:DRS RII:DRS RII:DRP RII:DRP SIGNATURE JHB /RA for/ Via email BRB /RA for/ ORL /RA for/ BRB /RA for/ BRB /RA for/ JDH /RA/ JHB /RA/ NAME GSmith WDeschaine MSpeck LLake RHamilton RKellner JHamman JBartley DATE 02/07/2014 02/07/2014 02/07/2014 02/07/2014 02/07/2014 02/07/2014 02/07/2014 02/07/2014 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO
J. Shea 3
Letter to J.W. Shea from Jonathan H. Bartley dated February 7, 2014
SUBJECT: SEQUOYAH NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT
05000327/2013005 AND 05000328/2013005
Distribution w/encl
- C. Evans, RII L. Douglas, RII
OE Mail
RIDSNRRDIRS
PUBLIC
RidsNrrPMSequoyah Resource
Enclosure U. S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos.: 50-327, 50-328
Report Nos.: 05000327/2013005, 05000328/2013005
Licensee: Tennessee Valley Authority (TVA)
Facility: Sequoyah Nuclear Plant, Units 1 and 2
Location: Sequoyah Access Road Soddy-Daisy, TN 37379
Dates: October 1 - December 31, 2013
Inspectors: G. Smith, Senior Resident Inspector W. Deschaine, Resident Inspector M. Speck, Senior Emergency Preparedness Inspector (Sections
1R04.1 and 1R05) L. Lake, Senior Reactor Inspector (Section 1R08) R. Hamilton, Senior Health Physicist (Section 2RS8)
R. Kellner, Health Physicist (Sections 2RS1, 4OA1)
Approved by: Jonathan H. Bartley, Chief Reactor Projects Branch 6 Division of Reactor Projects
Enclosure SUMMARY
IR 05000327/2013-005, 05000328/2013-005; 10/1/2013 - 12/31/2013; Sequoyah Nuclear Plant, Units 1 and 2; Other Activities
The report covered a three-month period of inspection by resident inspectors and announced
inspections by regional inspectors. One self-revealing finding was 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) dated June 2, 2011. Cross-cutting aspects are determined using IMC 0310, "Components Within the Cross-Cutting Areas,"
dated October 28, 2011. The NRC's program for ov
erseeing 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: A self-revealing non-cited violation of 10 CFR 50 Appendix B, Criterion XVI, "Corrective Action," was identified for the licensee's failure to promptly correct a
condition adverse to quality within a reasonable time. Timely corrective actions were not
taken to correct a dual position indication (open and closed lights both illuminated) on
the Unit 1 'A' train residual heat removal (RHR) containment sump suction flow control
valve (FCV) 1-FCV-63-72. This licensee entered this issue into the corrective action program as problem evaluation report (PER) 772193 and performed repairs to the valve to restore the system to operable status.
This finding was determined to be more than minor because it was associated with the
Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone's objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core
damage). Specifically, the finding reduced the reliability and capability of the 'A' train
RHR system to perform its safety function as designed. The finding required a detailed risk analysis as the 'A' RHR system was inoperable beyond its allowed outage time of 72
hours. The detailed risk analysis concluded that the finding was of very low safety significance (Green). This finding was determined to have a cross-cutting aspect relating to the proper classification, prioritization, and evaluation of operability and
reportability of conditions adverse to quality in the Corrective Action component of the
Problem Identification and Resolution area. P.1(c) (Section 4OA5)
B. Licensee-Identified Violations
None
Enclosure REPORT DETAILS
Summary of Plant Status
Unit 1 operated at or near 100 percent rated thermal power (RTP) until September 9, 2013,
when the unit entered a power coast down period until October 14 when the unit shut down for a
refueling outage. Unit 1 returned to 100 percent RTP on November 24 where it operated for the
remainder of the inspection period.
Unit 2 operated at or near 100 percent RTP for the entire inspection period.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection
a. Inspection Scope
.1 Readiness for Seasonal Extreme Weather Conditions
The inspectors reviewed design features and licensee preparations for protecting the
essential raw cooling water (ERCW) intake structure and both Unit 1 and 2 refueling
water storage tanks (RWSTs) from extreme cold and freezing conditions. The
inspectors reviewed the Updated Final Safe
ty Analysis Report (UFSAR) and Technical Specifications (TS), reviewed implementation of licensee freeze protection procedures,
walked down portions of the systems to assess deficiencies and system readiness for extreme cold weather, and discussed prioritization and status of correcting deficiencies
with licensee personnel. Documents reviewed are listed in the Attachment. The
inspectors completed one sample.
b. Findings
No findings were identified.
1R04 Equipment Alignment
.1 Partial System Walkdown
a. Inspection Scope The inspectors performed partial walkdowns of the following three 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
4 Enclosure 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. The inspectors completed 3 samples.
- Spent fuel pool cooling during core empty period of U1R19
- 1A emergency core cooling train while 1B 669 penetration cooler out-of-service
- 2A auxiliary feed-water and 2A emergency diesel generator while 2B under-voltage coils out-of-service
.2 Complete System Walkdown
a. Inspection Scope The inspectors performed a complete system walkdown of the: 1) emergency gas
treatment system/auxiliary building gas treat
ment system (ABGTS
); and 2) auxiliary building ventilation/control building ventilation systems. The purpose of this inspection
was 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 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:
- Dampers were correctly positioned.
- Electrical power was available as required.
- Hangers and supports were correctly installed and functional.
- Essential support systems were operational.
- Ancillary equipment or debris did not interfere with system performance.
- Breakers were correctly positioned.
- Major system components were correctly labeled.
- Cabinets, cable trays, and conduits were correctly installed and functional.
- Visible cabling appeared to be in good material condition.
In addition, the inspectors reviewed corrective action items and design issues associated
with the systems to determine whether any condition described in those documents
could adversely impact current system operability. Documents reviewed are listed in the Attachment. The inspectors completed two samples.
b. Findings
No findings were identified.
5 Enclosure 1R05 Fire Protection
.1 Fire Protection Tours
a. Inspection Scope
The inspectors conducted a tour of the six areas important to safety listed below to assess the material condition and operational status of fire protection features. The inspectors evaluated whether: combustibles and ignition sources were controlled in
accordance with the licensee's administrative procedures; fire detection and suppression
equipment was available for use; passive fire barriers were maintained in good material
condition; and compensatory measures for out-of-service, degraded, or inoperable fire protection equipment were implemented in accordance with the licensee's fire plan. Documents reviewed are listed in the Attachment. The inspectors completed six
samples.
- Unit 1 Lower Containment Building
- Unit 1 Upper Containment Building
- Control Building Elevation 685 (Auxiliary Instrument Room)
- Control Building Elevation 706 (Cable Spreading Room)
- ERCW Building - Elevations 688/704/720
- Turbine Building - Elevations 662/685
b. Findings
No findings were identified.
1R06 Flood Protection Measures
a. Inspection Scope
The inspectors examined internal flood protection measures associated with the 1A and 1B safety injection (SI) pump rooms' internal flood design in order to verify that flood mitigation plans were consistent with the design requirements and risk analysis
assumptions. The inspectors verified that equipment essential for reactor shutdown was
properly protected from a flood caused by pipe breaks in the 1A & 1B SI pump room. Specifically, the inspectors reviewed the licensee's moderate energy line break flooding study to fully understand the licensee's flood mitigation strategy, reviewed licensee drawings and then verified that the assumptions and results remained valid. The inspectors walked down the 1A & 1B SI pump room to verify the assumed flooding sources, adequacy of common area drainage, and flood detection instrumentation to
ensure that a flooding event would not impact reactor shutdown capabilities. The
inspectors completed one sample.
6 Enclosure b. Findings No findings were identified.
1R08 Non-Destructive Examination
Activities and Welding Activities
a. Inspection Scope
From October 21-25, 2013, the inspectors conducted an on-site review of the
implementation of the licensee's in-service inspection (ISI) Program for monitoring
degradation of the reactor coolant system; emergency feedwater systems, risk-
significant piping and components, and containment systems in Unit 1.
The inspectors' activities included a review of non-destructive examinations (NDEs) to
evaluate compliance with the applicable edition of the American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, and to verify that
indications and defects were appropriately evaluated and dispositioned in accordance with the requirements of the ASME Code,Section XI, acceptance standards or NRC
approved alternative requirement.
The inspectors directly observed or reviewed records of the following NDEs mandated
by the ASME Code to evaluate compliance with the ASME Code Section XI and Section
V requirements, and if any indications and defects were detected. Inspectors also reviewed evaluations of results that were dispositioned in accordance with the ASME Code or an NRC-approved alternative requirement.
- Directly observed:
o Ultrasonic testing (UT) examinations of the reactor pressure vessel head to shell
flange studs
o General visual examination of the outside surface of the containment shell
- Reviewed records:
o UT examinations of reactor coolant pump #4 bolting
o VT-3 visual examination of containment penetration bolting
o Work Order 113312025 modification of component cooling water system piping
The inspectors reviewed documentation for
the repair/replacement of the following pressure boundary welds. The inspectors evaluated if the licensee applied the pre-service non-destructive examinations and acceptance criteria required by the Construction Code. In addition, the inspectors reviewed the welding procedure
specifications, welder qualifications, welding material certifications, and supporting weld
procedure qualification records to evaluate if the weld procedures were qualified in
accordance with the requirements of the Construction Code and the ASME Code Section XI.
7 Enclosure PWR Vessel Upper Head Penetration (VUHP) Inspection Activities: For the Unit 1 vessel head, a bare metal visual examination and a volumetric examination required in accordance with the requirements of ASME Code Case N-729-1 and 10 CFR
50.55a(g)(6)(ii)(D) were conducted in the previous outage and therefore not required to
be performed this outage.
Boric Acid Corrosion Control (BACC) Inspection Activities: The inspectors reviewed the licensee's 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 on-site record review of procedures and the results of the
licensee's containment walkdown inspections performed during the current refueling outage. The inspectors also reviewed Focused Self-Assessment CRP-ENG-F-13-031 of
the Boric Acid Program.
The inspectors also interviewed the BACC program owner, conducted an independent
walkdown of containment to evaluate compliance with licensee's BACC program requirements, and verified that degraded or non-conforming conditions, such as boric acid leaks, were properly identified and corrected in accordance with the licensee's
BACC and corrective action programs.
The inspectors reviewed the following evaluations and 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.
- PER 691545 - Boric acid build up and wet boric acid are present on transmitter sensing line 1-FT-72-41
Steam Generator (SG) Tube Inspection Activities
- There were no SG tube eddy current examinations conducted during this outage. The inspectors reviewed the following documentation and evaluated them against the
licensee's TS, commitments made to the NRC, ASME Section XI, and Nuclear Energy
Institute (NEI) 97-06, "Steam Generator Program Guidelines," to ensure that the licensee
was in compliance with the schedule to skip the SG eddy current testing inspections for
the 1R19 outage:
- AREVA document # 51-9178898-001, Sequoyah Unit Condition Monitoring for Cycle 18 and Operational Assessment for Cycles 19, 20 and 21
Identification and Resolution of Problems
The inspectors performed a review of selected ISI-related problems that were identified by the licensee and entered into the corrective action program as PERs. The inspectors reviewed the PERs to confirm the licensee had appropriately described the scope of the
problem and had initiated corrective actions. The review also included the licensee's
8 Enclosure 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
No findings were identified.
1R11 Licensed Operator Requalification Program
.1 Quarterly Review of Licensed Operator Requalification
a. Inspection Scope
The inspectors performed one licensed operator requalification program review. The
inspectors observed a simulator session on October 9, 2013. The training scenario involved Just-In-Time Training for Pre-Refueling Outage risk significant activities such as placing the RHR system in service. The ins
pectors 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 TS action; and, group dynamics involved in crew performance. The inspectors also observed the evaluators' critique and reviewed
simulator fidelity to verify that it matched actual plant response. Documents reviewed
are listed in the Attachment. The inspectors completed one sample.
b. Findings
No findings were identified.
.2 Quarterly Review of Licensed Operator Performance
a. Inspection Scope
The inspectors observed and assessed licensed operator performance in the main
control room during periods of heightened activity or risk. The inspectors reviewed
various licensee policies and procedures such as OPDP-1, "Conduct of Operations," NPG-SPP-10.0, "Plant Operations," and 0-GO-5, "Normal Power Operation." The inspectors utilized activities such as post-maintenance testing, surveillance testing,
unplanned transients, infrequent plant evolutions, plant startups and shutdowns, reactor
power and turbine load changes, and refueling and other outage activities to focus on
the following conduct of operations as appropriate:
- operator compliance and use of procedures
- control board manipulations
- communication between crew members
9 Enclosure
- 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 1 reactor shutdown and plant cool down/depressurization
- Unit 1 refueling and other outage activities
- Unit 1 startup, including Mode changes
- Unit 2 down power with turbine in manual for valve testing
Documents reviewed are listed in the Attachment. The inspectors completed one
sample.
b. Findings
No findings were identified.
.3 Annual Review of Licensee Requalification Examination Results
a. Inspection Scope
On September 13, 2013, the licensee completed the annual requalification operating
examinations required to be administered to all licensed operators in accordance with 10
CFR 55.59(a)(2), "Requalification requirements," of the NRC's "Operators' Licenses."
The inspectors performed an in-office review
of the overall pass/fail results of the individual operating examinations and the crew simulator operating examinations in accordance with Inspection Procedure (IP) 71111.11, "Licensed Operator Requalification
Program and Licensed Operator Performance." The results were compared to the
thresholds established in Section 3.02, "Requalification Examination Results," of IP
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness
a. Inspection Scope
The inspectors reviewed five maintenance activities, issues, and/or systems listed below
to verify the effectiveness of the licensee's activities in terms of: appropriate work
practices; identifying and addressing common cause failures; scoping in accordance with 10 CFR 50.65(b); characterizing reliability issues for performance; trending key
10 Enclosure 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 structures, systems, 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. The
inspectors completed 5 samples.
- Cause Determination and Evaluation (CDE) #2700, FCV-63-72 Failure
- CDE #2696, EBGTS 'B' Fan Failure
- CDE #2686, 'A' Shutdown Boardroom Chiller Failure
- CDE #2674, 'B' Main Condenser Test Connection Failure
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 w
hether risk assessments were performed as required by 10 CFR 50.65(a)(4), and were accurate and complete. When emergent
work was performed, the inspectors reviewed whether plant risk was promptly
reassessed and managed. The inspectors also assessed whether the licensee's risk assessment tool use and risk categories were in accordance with Standard Programs and Processes Procedure NPG-SPP-07.1, "On-Line Work Management," Revision 3,
and Instruction 0-TI-DSM-000-007.1, "Risk Assessment Guidelines," Revision 9.
Documents reviewed are listed in the Attachment. The inspectors completed 2 samples.
- Review U1R19 Outage Schedule
- Review of risk during ABGTS outage
b. Findings
No findings were identified.
1R15 Operability Determinations and Functionality Assessments
a. Inspection Scope
For the eight 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
11 Enclosure evaluations to UFSAR descriptions to determine if the system or component's intended function(s) were adversely impacted. In addition, the inspectors reviewed compensatory measures implemented to determine whether the compensatory measures worked as stated and the measures were adequately controlled. The inspectors also reviewed a
sampling of PERs to assess whether the licensee was identifying and correcting any
deficiencies associated with operability evaluations. Documents reviewed are listed in
the Attachment. The inspectors completed 8 samples.
- PER 789552 - Unit 2 Turbine Controls in Manual
- PER 795451 - POE WO 113223153 T1 motor lead pinch
- PER 799097 - POE TS LCO 3.7.4 action for FCV-67-146
- PER 800432 - POE (ABSCE boundary issue)
- PER 795433 - PDO (During U1R19 water found leaking out of conduit in bioshield
wall) * PER 801415 - PDO EDG 1B 2 sec load sequence
- PER 803833 - PDO U-1 Rx Head Vent Valve Stroke
- PERs 816731, 815638, 817841 - FEs associated with the Unit 1 loose parts alarm
b. Findings
No findings were identified.
1R18 Plant Modifications
.1 Permanent Modifications
a. Inspection Scope
The inspectors reviewed the modification listed below and the associated 10 CFR 50.59
screening, and compared it against the UFSAR and TS to verify whether the
modification affected operability or availability of the affected system.
- DCN 22643 - Replace Pressurizer Power Operated Relief Valves (PORVs)
Following installation and testing, the inspectors observed indications affected by the modification, discussed them with operators, and verified that the modification was installed properly and its operation did not adversely affect safety system functions. The inspectors did note that, ultimately, the installed PORVs did not meet the acceptance
criteria associated with the close stroke time. As a result, the licensee chose to cut
out/remove the new style PORVs and reinstall the original PORVs prior to plant startup in November 2013. Documents reviewed are listed in the Attachment. The inspectors
completed one sample.
b. Findings
No findings were identified.
12 Enclosure 1R19 Post Maintenance Testing
a. Inspection Scope
The inspectors reviewed the post maintenance tests associated with the nine work
orders (WO) listed below to assess whether procedures and test activities ensured
system operability and functional capability. The inspectors reviewed the licensee's test procedure to evaluate whether: the procedure adequately tested the safety function(s) that may have been affected by the maintenance activity; the acceptance criteria in the
procedure were consistent with information in the applicable licensing basis and/or
design basis documents; and the procedure had been properly reviewed and approved. The inspectors also witnessed the test or reviewed the test data to determine whether test results adequately demonstrated restoration of the affected safety function(s). Documents reviewed are listed in the Attachment. The inspectors completed nine
samples.
- WO 113377829 - Repack Valve (1-LCV-3-175) and perform AIRCET test
- WO 112096045 - Repair isolation check valve (1-VLV-026-1296)
- WO 111234712 - 5 year PM to swap 480V Shutdown board breaker with a
refurbished breaker
- WO 113806636 - Perform 0-MI-EPM-317-102.0 on CCS pump C-S and
clean/replace motor air filter
- WO 114560807 - Centrifugal charging pump (CCP) room cooler fan motor current check, bearing lubrication and cleaning
- WO 114198329 - EQ maintenance and inspection
- WO 113408190 - Change out electrolytic capacitors in the Woodward 2301A
governor card
836 & 837
- WO 113756597 - PORVs - PCV-68-340 & PCV-68-334
b. Findings
No findings were identified.
1R20 Refueling and Other Outage Activities
.1 Unit 1 Refueling Outage Cycle 19
a. Inspection Scope
For the Unit 1 refueling outage that began on October 14, 2013, the inspectors
evaluated licensee activities 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 1 not normally accessible during at-power
13 Enclosure operations to verify that safety-related and risk-significant SSCs were maintained in an operable condition. Specifically, between October 14 and November 21, the inspectors performed inspections and reviews of the following outage activities. Documents
reviewed are listed in the Attachment. The inspectors completed one sample.
- 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) and
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 of the "B" Train CCP system to verify that the
equipment was appropriately configured to safely support the work and testing. To
ensure that RCS level instrumentation was properly installed and configured to give accurate information, the inspectors reviewed the installation of the Mansell level monitoring system. Specifically, the inspectors discussed the system with engineering, walked it down to verify that it was installed in accordance with
procedures and adequately protected from inadvertent damage, verified that Mansell
indication properly overlapped with pressurizer level instruments during pressurizer drain-down, verified that operators properly set level alarms to procedurally required set-points, and verified that the system consistently tracked RCS level while lowering
to reduced inventory conditions. The inspectors also observed operators compare
the Mansell indications with locally-installed ultrasonic level indicators during entry into reduced inventory conditions.
14 Enclosure
- Refueling Activities. The inspectors observed fuel movement at the spent fuel pool and at the refueling cavity in order to verify compliance with TS and that each assembly was properly tracked from core offload to core reload. In order to verify proper licensee control of foreign material, the inspectors verified that personnel were properly checked before entering any foreign material exclusion (FME) areas,
reviewed FME procedures, and verified that the licensee followed the procedures.
To ensure that fuel assemblies were loaded in the core locations specified by the
design, the inspectors independently reviewed the recording of the licensee's final
core verification.
- Reduced Inventory and Mid-Loop Conditions. Prior to the outage, the inspectors reviewed the licensee's commitments to Generic Letter 88-17. Before entering reduced inventory conditions the inspectors verified that these commitments were in
place, that plant configuration was in accordance with those commitments, and that
distractions from unexpected conditions or emergent work did not affect operator
ability to maintain the required reactor vessel level. Mid-loop conditions were not entered during this outage since SG eddy current testing was not required.
- Heat-up and Start-up Activities. The inspectors toured the containment prior to reactor startup to verify that debris that could affect the performance of the containment sump had not been left in the containment. The inspectors reviewed
the licensee's mode-change checklists to ve
rify that appropriate prerequisites were met prior to changing TS modes. To verify RCS integrity and containment integrity, the inspectors further reviewed the licensee's RCS leakage calculations and containment isolation valve lineups. In order to verify that core operating limit
parameters were consistent with core design, the inspectors also examined portions
of the low power physics testing surveillance.
b. Findings
No findings were identified.
1R22 Surveillance Testing
a. Inspection Scope
For the twelve
surveillance tests identified below, the inspectors assessed whether the SSCs involved in these tests satisfied the requirements described in the TS surveillance
requirements, the UFSAR, applicable licensee procedures, and whether the tests demonstrated that the SSCs were capable of performing their intended safety functions. This was accomplished by witnessing testing and/or reviewing the test data. Documents
reviewed are listed in the Attachment. The inspectors completed twelve
samples.
15 Enclosure
In-Service Tests
- * 1-SI-SXP-003-202.B, Motor Driven Auxiliary Feedwater Pump 1B-B Comprehensive Performance Test, Revision 7
- 1-SI-SXP-074-202.0, RHR Pump 1A-A and 1B-B Comprehensive Performance and Check Valve Test, Revision 10
RCS leakage test
- * 0-SI-OPS-068-137.0, Reactor Coolant System Water Inventory, Revision 32
Routine Surveillance Tests
- * 1-SI-OPS-088-001.0, Phase A Isolation Test, Revision 14
- 1-SI-OPS-082-026A, Loss of Offsite Power with Safety Injection D/G 1A-A Test, Revision 46
- 0-SI-SFT-072-138.0, Unit 1 Containment Spray - Spray Nozzle Test, Revision 6
- 0-SI-SXV-063-203.2, Full Stroking of Safety Injection Cold Leg Accumulator Isolation Valves, Revision 1
Ice Condenser Surveillance Test
- * 0-SI-MIN-061-105.0, Ice Condenser Ice Weighing, Revision 11
Containment Isolation Valve (CIV) Surveillance Tests
- * 0-SI-SLT-067-258.2, Containment Isolation Valve Local Leak Rate Test Lower Compartment Essential Raw Cooling Water, Revision 13
- 0-SI-SLT-088-259.4, Upper Personnel Airlock Interlock Operability Test, Revision 2
- 0-SI-SLT-088-259.1, Upper Personnel Airlock Overall Leak Rate Test and General Inspection, Revision 6
- 0-SI-SLT-081-258.1, Containment Isolation Valve Local Leak Rate Test Primary Water System, Revision 5
b. Findings
No findings were identified.
Cornerstone: Emergency Preparedness
1EP2 Alert and Notification System Evaluation
a. Inspection Scope
The inspectors evaluated the adequacy of the licensee's methods for testing and
maintaining the alert and notification sy
stem in accordance with NRC Inspection
Procedure 71114, Attachment 02, Alert and Notification System Evaluation. The
16 Enclosure applicable planning standard, 10 CFR Part 50.47(b)(5) and its related 10 CFR Part 50, Appendix E, Section IV.D requirements were used as reference criteria. The criteria contained in NUREG-0654, Criteria for Preparation and Evaluation of Radiological
Emergency Response Plans and Preparedness in Support of Nuclear Power Plants,
Revision 1, were also used as a reference.
The inspectors reviewed various documents which are listed in the Attachment, interviewed personnel responsible for system performance, and observed aspects of periodic siren maintenance and testing. This inspection activity satisfied one inspection
sample for the alert and notification system on a biennial basis.
b. Findings
No findings were identified.
1EP3 Emergency Response Organization Staffing and Augmentation System
a. Inspection Scope
The inspectors reviewed the licensee's Emergency Response Organization (ERO)
augmentation staffing requirements and process for notifying the ERO to ensure the
readiness of key staff for responding to an event and timely facility activation. The
qualification records of key position ERO personnel were reviewed to ensure all ERO qualifications were current. A sample of problems identified from augmentation drills or system tests performed since the last inspection was reviewed to assess the
effectiveness of corrective actions.
The inspection was conducted in accor
dance with NRC Inspection Procedure 71114, Attachment 03, Emergency Response Organization Staffing and Augmentation System. The applicable planning standard, 10 CFR 50.47(b)(2), and its related 10 CFR 50,
Appendix E requirements were used as reference criteria.
The inspectors reviewed various documents which are listed in the Attachment. This
inspection activity satisfied one inspection sample for the ERO staffing and augmentation system on a biennial basis.
b. Findings
No findings were identified.
1EP4 Emergency Action Level and Emergency Plan Changes
a. Inspection Scope
The NRC Office of Nuclear Security and Incident Response headquarters staff performed an in-office review of the latest revisions of various Emergency Plan Implementing Procedures (EPIPs) and the Emergency Plan located under ADAMS
17 Enclosure Accession numbers ML12326A678, ML12353A050, ML13025A102, ML13070A025, ML13219A022, and ML13246A091.
The licensee determined that in accordance with 10 CFR 50.54(q), the changes made in
the revisions resulted in no reduction in the effectiveness of the Plan, and that the
revised Plan continued to meet the requirements of 10 CFR 50.47(b) and Appendix E to
10 CFR Part 50. The NRC review was not documented in a safety evaluation report and did not constitute approval of licensee-generated changes; therefore, these revisions are subject to future inspection. Documents reviewed are listed in the Attachment. The
inspectors completed one sample.
b. Findings
No findings were identified.
1EP5 Maintenance of Emergency Preparedness
a. Inspection Scope
The inspectors reviewed the corrective actions identified through the Emergency
Preparedness program to determine the significance of the issues, the completeness
and effectiveness of corrective actions, and to determine if issues were recurring. The
licensee's post-event after action reports, self-assessments, and audits were reviewed to assess the licensee's ability to be self-critical, thus avoiding complacency and degradation of their emergency preparedness program. Inspectors reviewed the
licensee's 10 CFR 50.54(q) change process, personnel training, and selected
screenings and evaluations to assess adequacy. The inspectors toured facilities and
reviewed equipment and facility maintenance records to assess licensee's adequacy in maintaining them. The inspectors evaluated the capabilities of selected radiation monitoring instrumentation to adequately support Emergency Action Level (EAL)
declarations.
The inspection was conducted in accordanc
e with NRC Inspection Procedure 71114.05, Maintenance of Emergency Preparedness. The applicable planning standards, related 10 CFR 50, Appendix E requirements, and 10 CFR 50.54(q) and (t) were used as reference criteria.
The inspectors reviewed various documents which are listed in the Attachment. This
inspection activity satisfied one inspection sample for the maintenance of emergency preparedness on a biennial basis.
b. Findings
No findings were identified.
18 Enclosure 2. RADIATION SAFETY (RS)
Cornerstones: Occupational Radiation Safety (OS) and Public Radiation Safety (PS)
2RS1 Radiological Hazard Assessment and Exposure Controls
a. Inspection Scope
Hazard Assessment and Instructions to Workers: During facility tours, the inspectors directly observed labeling of radioactive material and postings for radiation areas, high radiation areas (HRAs), and airborne radioactivity areas established within the
radiologically controlled area (RCA) of the Unit 1 containment, Unit 1 and Unit 2 auxiliary buildings, Independent Spent Fuel Storage Installation (ISFSI), and radioactive waste (radwaste) processing and storage locations. The inspectors independently measured
radiation dose rates or directly observed conduct of licensee radiation surveys for RCA
areas in the Unit 1 containment, Unit 1 and Unit 2 Auxiliary buildings, and ISFSI. The
inspectors reviewed survey records for several plant areas including surveys for alpha emitters, airborne radioactivity, and pre-job surveys for selected Unit 1 Refueling Outage 19 (U1R19) tasks. The inspectors also discussed changes to plant operations that could
contribute to changing radiological conditions since the last inspection and reviewed
U1R19 crud burst results and post crud burst dose rate surveys. For selected U1R19
outage jobs, the inspectors attended, or reviewed, pre-job briefings and radiation work
permit (RWP) details to assess communication of radiological control requirements and current radiological conditions to workers. Selected U1R19 work activities included Unit 1 control rod drive mechanism duct work, Unit 1 Refueling Activities, Unit 1 Head O-ring
Surface Work & Inspection, and work in the Unit 1 Equipment Pit and transfer canal.
Hazard Control and Work Practices: The inspectors evaluated access barrier effectiveness for selected Unit 1 and Unit 2 Locked High Radiation Area (LHRA) and Very High Radiation Area (VHRA) locations. Changes to procedural guidance for LHRA
and VHRA controls were discussed with health physics (HP) supervisors. Controls and
their implementation for storage of irradiated material within the spent fuel pool (SFP)
were reviewed and discussed in detail. Established radiological controls (including
airborne controls) were evaluated for selected U1R19 tasks including refueling and reactor cavity work activities, work in auxiliary building HRAs, and radwaste 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.
Occupational workers' adherence to selected RWPs and HP technician (HPT)
proficiency in providing job coverage were evaluated through direct observations and
interviews with licensee staff. Electronic dosimeter (ED) alarm set points and worker
stay times were evaluated against area radiation survey results for refueling and reactor cavity work. ED alarm logs were reviewed and worker response to dose and dose rate
alarms during selected work activities was evaluated. For HRA tasks involving significant dose rate gradients, e.g. reactor head O-ring work, the inspectors evaluated the use and placement of whole body and extremity dosimetry to monitor worker
exposure.
19 Enclosure Control of Radioactive Material: The inspectors observed surveys of material and personnel being released from the RCA using small article monitor, personnel
contamination monitor, and portal monitor instruments. The inspectors reviewed the last two calibration records for selected release point survey instruments and discussed
equipment sensitivity, alarm set points, and release program guidance with licensee
staff. The inspectors compared recent 10 CFR Part 61 results for the Dry Active Waste
(DAW) radioactive waste stream with radionuclides used in calibration sources to evaluate the appropriateness and accuracy of
release survey instrumentation. The inspectors also reviewed records of leak tests on selected sealed sources and discussed
nationally tracked source transactions with licensee staff.
Problem Identification and Resolution: PERs associated with radiological hazard assessment and control were reviewed and assessed. The inspectors evaluated the licensee's ability to identify and resolve the issues in accordance with procedure NPG-
SPP-22.300, "Corrective Action Program," Rev. 0. The inspectors also evaluated the
scope of the licensee's internal audit program and reviewed recent assessment results.
Radiation protection activities were evaluated against the requirements of UFSAR Section 12; TS Sections 6.8 and 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. The inspectors completed one sample.
b. Findings
No findings were identified.
2RS8 Radioactive Solid Waste Processing and Radioactive Material Handling, Storage, and
Transportation
a. Inspection Scope
Waste Processing and Characterization: During inspector walkdowns, accessible sections of the liquid and solid radwaste processing systems were assessed for material condition and conformance with system design diagrams. Inspected equipment included radwaste storage tanks; resin transfer piping, resin, and filter packaging components;
and abandoned boric acid evaporator equipment. The inspectors discussed component function, processing system changes, and radwaste program implementation with licensee staff.
The radionuclide characterizations for 2010, and 2012, for selected waste streams were reviewed and discussed with Radwaste/Transportation staff. For primary resin, reactor
coolant system filters, and DAW, the inspectors evaluated analyses for hard-to-detect
nuclides, reviewed the use of scaling factors, and examined quality assurance comparison results between licensee waste stream characterizations and outside laboratory data. Waste stream mixing and concentration averaging methodology for
resins and filters was evaluated and discussed with Radwaste/Transportation staff. The
20 Enclosure inspectors also reviewed the licensee's procedural guidance for monitoring changes in waste stream isotopic mixtures. The 10 CFR 61 analysis results were also discussed
with Chemistry personnel.
Radioactive Material Storage: During walkdowns of indoor and outdoor radioactive material storage areas, the inspectors observed the physical condition and labeling of
storage containers and the posting of Radioactive Material Areas. The inspectors also reviewed licensee procedural guidance for storage and monitoring of radioactive
material.
Transportation: The inspectors observed a shipment of vendor equipment during the week of inspection. The inspectors reviewed shipping procedure requirements and discussed preparation of shipping documents, package marking and labeling, and interviewed shipping technicians regarding Department of Transportation (DOT)
regulations.
Selected shipping records were reviewed for consistency with licensee procedures and compliance with NRC and DOT regulations. The inspectors reviewed emergency response information, DOT shipping package classification, waste classification, radiation survey results, and evaluated whether receiving licensees were authorized to
accept the packages. Licensee procedures for handling shipping containers were
compared to Certificate of Compliance requirements and manufacturer
recommendations. In addition, training records for selected individuals currently qualified to ship radioactive material were reviewed.
Radwaste processing activities and equipment configuration were reviewed for
compliance with the licensee's Process Control Program and UFSAR, Chapter 11.
Waste stream characterization analyses were reviewed against regulations detailed in 10 CFR Part 20, 10 CFR Part 61, and guidance provided in the Branch Technical Position on Waste Classification (1983). Radioactive material and waste storage
activities were reviewed against the requirements of 10 CFR Part 20. Transportation
program implementation was reviewed against regulations detailed in 10 CFR Part 20,
10 CFR Part 71, 49 CFR Parts 172-178, as well as the guidance provided in NUREG-
1608. Training activities were assessed against 49 CFR Part 172, Subpart H.
Problem Identification and Resolution: The inspectors reviewed PERs in the area of radwaste processing and transportation. The inspectors evaluated the licensee's ability
to identify and resolve the issues in accordance with procedure NPG-SPP-22.300,. The
inspectors also evaluated the scope of the licensee's internal audit program and reviewed recent assessment results. Documents reviewed are listed in the Attachment. The inspectors completed one sample.
b. Findings
No findings were identified.
21 Enclosure 4. OTHER ACTIVITIES
4OA1 Performance Indicator (PI) Verification
a. Inspection Scope
Occupational Radiation Safety Cornerstone: The inspectors reviewed the Occupational Exposure Control Effectiveness PI results for the Occupational Radiation Safety Cornerstone from October 2012 through October 2013. For the assessment period, the
inspectors reviewed ED alarm logs and selected PERs related to controls for exposure
significant areas. The inspectors also reviewed licensee procedural guidance for
collecting and documenting PI data. Documents reviewed are listed in the Attachment. The inspectors completed one sample.
Emergency Preparedness Cornerstone
- * Drill/Exercise Performance (DEP)
- Emergency Response Organization Drill Participation (ERO)
- Alert and Notification System Reliability (ANS)
For the specified review period, the inspectors examined data reported to the NRC,
procedural guidance for reporting PI information, and records used by the licensee to
identify potential PI occurrences. The inspectors verified the accuracy of the PI for ERO drill and exercise performance through review of a sample of drill and event records. The inspectors reviewed selected training records to verify the accuracy of the PI for
ERO drill participation for personnel assigned to key positions in the ERO. The
inspectors verified the accuracy of the PI for alert and notification system reliability
through review of a sample of the licensee's records of periodic system tests. The
inspectors also interviewed the licensee personnel who were responsible for collecting and evaluating the PI data. Documents reviewed are listed in the Attachment. This inspection satisfied three inspection samples for PI verification on an annual basis.
b. Findings
No findings were identified.
4OA2 Problem Identification and Resolution
.1 Routine Review
a. Inspection Scope
As required by IP 71152, "Problem Identification and Resolution," and in order to help
identify repetitive equipment failures or specific human performance issues for follow-up,
the inspectors performed a daily screening of items entered into the licensee's CAP. This was accomplished by reviewing the description of each new PER and attending daily management review committee meetings.
22 Enclosure b. Findings
No findings were identified.
.2 Annual Follow-up of Selected Issues
a. Inspection Scope The inspectors performed an in-depth review of PER 665633, "NRC identified freeze
protection issues." The inspectors reviewed the actions taken to determine if the
licensee had adequately addressed the following attributes. Documents reviewed are
listed in the Attachment. The inspectors completed one sample for Annual Follow-up of
Selected Issues.
- 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
No findings were identified.
.3 Semiannual Trend Review
a. Inspection Scope
As required by IP 71152, the inspectors performed a review of the licensee's corrective action program and associated documents to identify trends that could indicate the
existence of a more significant safety issue. The inspectors review was focused on
repetitive equipment issues, but also included licensee trending efforts and licensee human performance results. The inspectors review nominally considered the twelve-
month period of January 2013 through December 2013, 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. Documents reviewed are listed in the Attachment. The inspectors completed one
sample for Semiannual Trend Review.
b. Findings and Observations
No findings were identified. In general, the licensee had identified trends and
23 Enclosure appropriately addressed them in their CAP. The inspectors evaluated the licensee trending methodology and observed that the licensee had performed a detailed review. The licensee routinely reviewed cause codes, involved organizations, key words, and
system links to identify potential trends in their data. The inspectors compared the
licensee process results with the results of the inspectors' daily screening. No
previously unidentified trends of significance were identified.
.4 Annual Follow-up of Operator Workarounds
a. Inspection Scope
The inspectors reviewed the operator workaround (OWA) program to verify that OWAs were identified at an appropriate threshold, were entered into the CAP, and that corrective actions were appropriate and timely. Specifically, the inspectors reviewed the licensee's workaround lists and repair schedules, reviewed CAP word searches,
conducted tours and interviewed operators and operations department support staff.
Additionally, the inspectors checked for undocumented workarounds by observing operators perform rounds, reviewed operator deficiency lists, reviewed appropriate system health documents, attended plant health committee meetings, and verified that
identified program deficiencies were corrected. The inspectors evaluated all
workarounds for their aggregate impact. Documents reviewed are listed in the
Attachment. The inspectors completed one sample for Annual Follow-up of Operator
Workarounds.
b. Findings
No findings were identified.
4OA5 Other Activities
.1 (Closed) Unresolved Item (URI) 050000327/2013004-01, Water Intrusion into Actuator of
Valve 1-FCV-63-72
a. Inspection Scope
The inspectors opened this URI as a result of water intrusion into the actuator of 1-FCV-
63-72, which is the 'A' train containment sump suction for the Unit 1 'A' RHR train. This issue was noted during an operability inspection conducted last quarter. The inspectors
determined more inspection was required in order to resolve the issue. On August 8, 2013, an operator noted the valve exhibited dual indication and on August 14, a related valve, 1-FCV-74-3, failed its periodic stroke test. The following day, 1-FCV-63-72 was
noted to be failed as well due to a large of amount of water buildup in the actuator. A
subsequent root cause of the failure was completed during this inspection period and
concluded the water intrusion was due to groundwater which migrated through the wall
of the RHR valve vault room and into the valve conduit. Although the circumstances regarding the water intrusion may have been beyond the licensee's ability to predict, the
24 Enclosure inspectors noted there were opportunities before August 14 to identify and correct the deficient condition. Thus, the inspectors identified the following non-cited violation (NCV) as discussed below. Documents reviewed are listed in the Attachment.
b. Findings
Introduction: A self-revealing Green NCV of 10 CFR 50 Appendix B, Criterion XVI, "Corrective Action," was identified for the licensee's failure to correct a condition adverse to quality within a reasonable amount of time. Timely corrective actions were not taken
to correct a dual position indication (open and closed lights both illuminated) on the Unit
1 'A' train RHR containment sump suction flow control valve 1-FCV-63-72.
Description: On August 8 at 0709, the Unit 1 control room operator noted that valve 1-FCV-63-72 showed dual position indication on the control board. This valve is the 'A'
train RHR suction valve from the reactor containment sump and is normally closed,
showing only a single position indication lamp on the control board. Valve 1-FCV-63-72
was verified to be locally closed. No other activities were noted that would have caused the valve to come off its closed seat. Initial troubleshooting for the dual indication consisted of: 1) a visual inspection of the valve; 2) a visual inspection of the motor control center (MCC) cubicle during an attempted closure of the valve; 3) a review of the
wiring diagram by a troubleshooting team; 4) replacement of the MCC light indicating
bulb; and 5) a visual inspection of the main control room (MCR) hand switch. Based on the troubleshooting team's analysis of the wiring diagrams, no impact was expected on the interlocks associated with 1-FCV-63-72. The team initially concluded that the most
likely cause of the indication was a short circuit in the control power indication in the
MCR valve hand switch. Based on this conclusion, plus the fact that the valve is not normally stroked at power (due to concerns of accidently transferring borated water from the RWST to the containment sump), the licensee chose not to immediately stroke test 1-FCV-63-72. Instead, the licensee declared the position indication for the valve
inoperable per Post Accident Monitoring requirements as delineated in TS 3.9.1. This was a 30 day limiting condition for operation. The licensee then began development of a
troubleshooting plan which would require more intrusive troubleshooting of the issue
starting the following week.
On August 14 at 2315, during a routine quarterly inservice testing valve stroke activity,
valve 1-FCV-74-3 failed to stroke in the closed direction from the control room. This
valve is the 'A' train RHR suction valve from the RWST and is normally open. Valve 1-FCV-74-3 was immediately declared out of service and the 72-hour Emergency Core Cooling Systems (ECCS) TS 3.5.2 action statement was entered. During troubleshooting, operators attempted to close valve 1-FCV-74-3 remotely from the MCC
cubicle. This action blew control power fuses. The licensee then attempted local
manual operation and noted 1-FVC-74-3 could be manually closed without binding.
Valve 1-FCV-74-3 was partially manually closed and then reopened from the MCC without incident. Due to the relationship between valves 1-FCV-63-72 and 1-FCV-74-3 (interlocks, shared wiring in junction boxes, etc.) the licensee suspected that the failure of valve 1-FCV-74-3 to close and valve 1-FCV-63-72 dual position indication were
related.
25 Enclosure The licensee subsequently opened the 1-FCV-63-72 actuator and noted that a significant amount of water had accumulated inside the actuator. This water caused significant electrical shorting in the valve control circuit and rendered the valve
inoperable. Also, the water affected valve 1-FCV-74-3, as this valve utilizes contacts
from valve 1-FCV-63-72 circuitry. It was noted that a low current short caused the failure
of the closing coil for valve 1-FCV-74-3. Following repairs to both 1-FCV-63-72 and
1-FCV-74-3, the ECCS system was returned to operable status on August 17 at 0200.
The licensee's past operability determination concluded that 1-FCV-63-72 and 1-FCV-
74-3 were likely inoperable beginning on August 8 when 1-FCV-63-72 was noted to have
a dual indication. Thus the 'A' train ECCS system was most likely inoperable for
approximately nine days, which exceeded the TS allowable outage time. On October 21, 2013, Licensee Event Report 50-327/2013-003 was submitted as a result of this issue. The licensee concluded that the source of the water was ground water that had
migrated through the concrete ceiling that housed the valve and actuator cables. The
ground water leaked through the threaded penetration seal and inside the conduit and
flowed down into the valve actuator. During the most recent Unit 1 refueling outage in November 2013, the licensee redesigned the conduit penetration to prevent the intrusion of moisture into the conduit. The licensee noted the rate of moisture intrusion was most
likely higher in the recent months due to a higher than normal amount of rainfall that
temporarily raised the water table in the vicinity of the plant. The inspectors also noted
that on February 29, 2012, the licensee discovered water buildup in the actuator of 1-
FCV-63-72. This deficiency was entered into the CAP; however it appears that this precursor was not adequately evaluated such that continued water intrusion ultimately led to the failure noted on August 8, 2013.
Analysis: The licensee's failure to take timely actions to correct a condition adverse to quality was a performance deficiency. The inspectors concluded that testing and inspection could have determined that valve 1-FCV-63-72 was inoperable much earlier than August 14 when it was noted that RHR suction valve to the RWST, 1-FCV-74-3, did
not pass its routine surveillance test. This finding was determined to be more than minor
because it was associated with the Design Control attribute of the Mitigating Systems
cornerstone and adversely affected the cornerstone's objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the finding reduced the reliability and capability of the 'A' train RHR system to perform its safety function as designed. Using IMC 0609.04, "Initial Characterization of Findings," dated June 19,
2012, and IMC 0609, Appendix A, Exhibit 4 - "External Events Screening Questions,"
dated June 19, 2012, the finding required a detailed risk analysis as the 'A' RHR system was inoperable beyond its TS-allowed outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The detailed risk analysis concluded that the finding was of very low safety significance (Green).
A Phase 3 analysis was performed by the regional Senior Reactor Analyst to determine the impact of the finding. The analysis assumed a recoverable failure of the 1-FCV-63-
72 valve, along with a dependent failure of the 1-FCV-74-3 valve. The major impacts were in the swapover from the RWST to the containment sump as the source of water to
26 Enclosure mitigate medium and smaller LOCA sequences. Because of the low exposure time, the availability of the opposite train, and the ability of the operations staff to operate the
effected valves manually, the finding was determined to be Green.
The cause of this finding was determined to have a cross-cutting aspect relating to the
proper classification, prioritization, and evaluation of operability and reportability of
conditions adverse to quality in the Corrective Action component of the Problem Identification and Resolution area, in that, on February 29, 2012, the licensee discovered water buildup in the actuator of 1-FCV-63-72 and did not adequately evaluated the
condition adverse to quality such that continued water intrusion ultimately led to the
failure noted on August 8, 2013. P.1(c)
Enforcement: Title 10 of the Code of Federal Regulations Part 50, Appendix B, Criterion XVI, "Corrective Action," requires, in part, that measures shall be established to assure
that conditions adverse to quality, such as failures, malfunctions, deficiencies,
deviations, defective material and equipment, and non-conformances are promptly
identified and corrected. Contrary to the above, from August 8 through August 17, 2013, the licensee failed to assure that a condition adverse to quality, the failure of valve FCV-63-72, was corrected in a timely manner. Specifically, the licensee failed to sufficiently
evaluate and correct a moisture intrusion problem associated with the RHR containment
suction motor-operated valve. Corrective actions taken by the licensee included
redesigning and modifying the conduit penetration to prevent the intrusion of moisture
into the conduit. The violation was entered into the licensee's CAP as PER 772193. This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy and will be identified as NCV 05000327/2013005-01, "Unit 1 Train
'A' RHR Containment Suction Valve Failure."
.2 Quarterly Resident Inspector Observations of Security Personnel and Activities
a. Inspection Scope
During the inspection period, the inspectors conducted observations of security force
personnel and activities to ensure that the activities were consistent with licensee security procedures and regulatory requirements relating to nuclear plant security. These observations took place during both normal and off-normal plant working hours.
These quarterly resident inspector observations of security force personnel and activities
did not constitute any additional inspection samples. Rather, they were considered an
integral part of the inspectors' normal plant status review and inspection activities.
b. Findings
No findings were identified.
27 Enclosure .3 Review of the Operation of an Independent Spent Fuel Storage Installation (ISFSI)
(60855.1) a. Inspection Scope
The inspectors performed a walkdown with the field operator of the ISFSI storage pad on
December 26, 2013, to verify that operations were conducted in a safe manner in accordance with approved procedures and without undue risk to the health and safety of the public. The inspectors noted that there were 40 multi-purpose canisters (MPCs)
positioned on the ISFSI pad. The inspectors verified the MPC vents were in good
condition and free of obstruction. The inspectors also verified natural circulation within
the MPCs. The inspectors verified that any ISFSI problems were placed in the CAP. The inspectors also reviewed ISFSI document control practices to verify that changes to the required ISFSI procedures and equipment were performed in accordance with
guidelines established in local procedures and 10 CFR 72.48. Documents reviewed are
listed in the Attachment.
b. Findings
No findings were identified.
4OA6 Meetings
.1 Exit Meeting Summary
On January 13, 2014, the resident inspectors presented the inspection results to Mr. Carlin and other members of his staff, who acknowledged the finding. The inspectors
asked the licensee whether any of the material examined during the inspection should be considered proprietary. No proprietary information was identified.
ATTACHMENT: SUPPLEMENTARY INFORMATION
Attachment SUPPLEMENTARY INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
J. Alfultis, Director of Modifications & Projects
J. Carlin, Site Vice President
J. Cross, Chemistry Manager A. Day, Radiation Protection Manager
D. Erb, Work Control Manager
M. Henderson, ISI Program Engineer
H. Hill, Rad Waste Superintendent
J. Johnson, Program Manager Licensing A. Little, Site Security Manager K. Loomis, Boric Acid Program Engineer
T. Marshall, Operations Manager
M. McBrearty, Licensing Manager
S. McCamy, Quality Assurance Manager S. Mohorn, Rad Waste Superintendent
P. Noe, Director Safety and Licensing
C. Owens, Rad Waste HP
W. Pierce, Site Engineering Director
P. Pratt, Manager, Maintenance
J. Rolph, Radiation Protection Technical Support Superintendent
P. Simmons, Plant Manager K. Smith, Director of Training
D. Sutton, Licensing
J. Stamey, Rad Waste Health Physicist
J. Stewart, Chemist
NRC personnel
S. Lingam, Project Manager, Office of Nuclear Reactor Regulation
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
05000327/2013005-01 NCV Unit 1 Train 'A' RHR Containment Suction Valve Failure (Section 4OA5)
Closed 05000327/2013004-01 URI Water Intrusion Into Actuator of Valve 1-
FCV-63-72 (Section 4OA5)
Enclosure LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection Procedures
0-PI-OPS-006.0, Freeze Protection, Rev. 55
Service Requests (SRs)
SR 807550
SR 825408
SR 821489
Section 1R04: Equipment Alignment
Partial System Walkdowns
Procedures
0-GO-16, System Operability Checklists, Rev. 4
Other documents
UFSAR Section 9
Procedures
0-SI-OPS-030-021.A, Auxiliary Building Gas Treatment System Train A, Rev. 6
0-SI-OPS-030-021.B, Auxiliary Building Gas Treatment System Train B, Rev. 6
0-SO-30-18, Auxiliary Building Gas Treatment System, Rev. 14 0-SO-65-1, Emergency Gas Treatment System Air Cleanup and Annulus Vacuum, Rev. 27 0-SO-30-1, Control Building Heating, Air Conditioning, and Ventilation, Rev. 39
0-SO-30-10, Auxiliary Building Ventilation Systems, Rev. 54
Section 1R05: Fire Protection
Procedures
FPDP-1, Conduct of Fire Protection, Rev. 2
0-PI-FPU-317-299.W, Att. 8, Shift Check List, Rev. 32
NPG-SPP-18.4.7, Control of Transient Combustibles, Rev. 0
EITP-100, Environmental Compliance, Rev. 6
0-SI-FPU-410-703.0, Inspection of FPR Required Fire Doors, Rev. 5 SQN-FPR-Part-II, SQN Fire Protection Report Part II - Fire Protection Plan, Rev. 28
Other documents
Fire Protection Pre-Fire Plans for Unit 1 Lower Containment Building
Fire Protection Pre-Fire Plans for Unit 2 Lower Containment Building Fire Protection Pre-Fire Plans for Control Building Elevation 685 (Auxilia
ry Instrument Room)
Fire Protection Pre-Fire Plans for Control Building Elevation 706 (Cable Spreading Room) Fire Protection Pre-Fire Plans for ERCW Building - Elevations 688/704/720
Fire Protection Pre-Fire Plans for Turbine Building - Elevations 662/685
Section 1R06: Flood Protection Measures
Work Orders
WO 11108121224, Check Standing Water Level in Manholes/Handholes
3 Attachment
Other documents
TVA letter to NRC dated May 4, 2007. TVA response to GL 2007-01
Section 1R08: Inservice Inspection Activities
Procedures
N-VT-15 - Visual Examination of Class MC and Metallic Liners of Class CC Components of
Light-Water cooled Plants, Rev. 11 N-VT-16 - General Visual Examination Containment Vessel Integrity Verification, Rev. 05 N-UT-67 - Generic Procedure for Straight Beam Ultrasonic Examination of Bolts and Studs,
Rev. 05
PDI-UT-5 - Generic Procedure for Straight Beam Ultrasonic Examination of Bolts and Studs,
Rev. D34 IEP-200 - Qualification and Certification Requirements for TVA Inspection Services Organization (ISO) Nondestructive (NDE) Personnel, Rev. 13
Corrective Action Documents
PER 618770 - Boron buildup on 1B-B SIS Pump Pedestal PER 691545 - Boric acid build up and wet boric acid are present on transmitter sensing line
1-ft-72-41
PER 01-010244 - Minor concrete voids in U1C11 Vt-3 inspection
PER 169175 - Airline cracks in ceiling beneath reactor cavity and reactor wall
SR 797854 - Hairline cracking in the concrete beneath the fuel transfer canal in lower
containment SR 526607 - Spalling on baseplate of Protection Device No. 1 on Drawing 48N1701-17. SR 797166 - Boric acid on Reactor Coolant Pump #1 on #3 seal
SR 797061 - Boric acid on valve 1-FCV-063-0098
SR 797072 - Two areas of white deposit in Fan Room 2
Other documents
Periodic Instruction 0-PI-DXI-000-116.2, ASME Section XI IWE/IWL Containment Inservice
Inspection (CSI) Program, Rev. 05
Q-NIC-100 - Written Practice for the Qualification and Certification of Nondestructive
Examination (NDE0 Personnel, Rev. 20-TVA
IHI Southwest Technologies, Inc. Operating Procedure 2.0-NDES-001, Nondestructive Examination Personnel Qualification and Certification, Rev. 06 WO 113312025 - Modify Component Cooling Piping to eliminate interference with actuator for
1-FC-063-011
Section 1R12: Maintenance Effectiveness
Procedures
TI-4, Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting -
10CFR50.65, Rev. 23
Other documents
MR 11th Periodic Assessment Report (PE sample) Cause Determination and Evaluation (CDE) #2700, FCV-63-72 Failure CDE #2696, EBGTS 'B' Fan Failure
4 Attachment CDE #2686, 'A' Shutdown Boardroom Chiller Failure CDE #2674, 'B' Main Condenser Test Connection Failure
Section 1R13: Maintenance Risk Assessments and Emergent Work Control
Procedures
0-TI-DSM-000-007.1, Risk Assessment Guidelines, Rev. 9
NPG-SPP-07.3, Work Activity Risk Management Process, Rev. 3 NPG-SPP-07.2.4, Forced Outage or Short Duration Planned Outage Management, Rev. 0 NPG-SPP-07.2, Outage Management, Rev. 0
GOI-6, Apparatus Operations, Rev. 142
Section 1R15: Operability Determinations and Functionality Assessments
Procedures
NEDP-22, Functional Evaluations, Rev. 9
OPDP-8, Limiting Conditions for Operation Tracking, Rev. 5
NPG-SPP-03.5, Regulatory Reporting Requirements, Rev. 2
PERs 789552 - Unit 2 Turbine Controls in Manual
795451 - POE WO 113223153 T1 motor lead pinch
799097 - POE TS LCO 3.7.4 action for FCV-67-146
800432 - POE (ABSCE boundary issue)
795433 - PDO (During U1R19 water found leaking out of conduit in bioshield wall) 801415 - PDO EDG 1B 2 sec load sequence 803833 - PDO U-1 Rx Head Vent Valve Stroke
816731, 815638, 817841 - FEs associated with the Unit 1 loose parts alarm
Section 1R18: Plant Modifications
Procedures
NPG-SPP-09.3, Plant Modifications and Engineering Change Control, Rev. 4
NPG-SPP-09.4, 10 CFR 50.59 Evaluations of Changes, Tests, and Experiments, Rev. 1
NPG-SPP-09.5, Temporary Alterations, Rev. 0
Other documents
DCN 22643 - Replace Pressurizer PORVs
Section 1R19: 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. 0
NPG-SPP-6.1, Work Order Process Initiation, Rev. 0
NPG-SPP-06.3, Pre-/Post-Maintenance Testing, Rev. 0
NPG-SPP-06.9, Testing Programs, Rev. 0
NPG-SPP-06.9.1, Conduct of Testing, Rev. 1 NPG-SPP-06.9.3, Post-Modification Testing, Rev. 0
5 Attachment
Work Orders
114306842 - Disassemble and reassemble valve in support of 113716425 114306841 - Remove actuator, install actuator, set up calibration in support of 113716425
114325805 - Disassemble and reassemble valve in support of 113716459
114325799 - Remove and install actuator in support of 113716459
113756597 - PORVs - PCV-68-340 & pcv-68-334 Replacement activities
113377829 - Repack Valve (1-LCV-3-175) and perform AIRCET test 112096045 - Repair isolation check valve (1-VLV-026-1296) 111234712 - 5 year PM to swap 480V Shutdown board breaker with a refurbished breaker
113806636 - Perform 0-MI-EPM-317-102.0 on CCS pump C-S and clean/replace motor air filter
114560807 - CCP room cooler fan motor current check, bearing lubrication and cleaning
114198329 - EQ maintenance and inspection 113408190 - Change out electrolytic capacitors in the Woodward 2301A governor card
Section 1R20: Refueling and Other Outage Activities
Procedures
FHI-3, Movement of Fuel, Rev. 65
0-GO-15, Containment Closure Control, Rev. 34 0-GO-13, Reactor Coolant System Drain and Fill Operations, Rev. 71
NPG-SPP-08.1, Nuclear Fuel Management, Rev. 00
0-PI-OPS-000-011.0, Containment Access Control During Modes 1-4, Rev. 1
Section 1R22: Surveillance Testing
Procedures
NPG-SPP-06.9.1, Conduct of Testing, Rev. 8
0-SI-SXV-072-266.0, ASME Code Valve Testing, Rev. 12
0-SI-OPS-068-137.0, Reactor Coolant System Water Inventory, Rev. 32
0-SI-SFT-072-138.0, Unit 1 Containment Spray - Spray Nozzle Test, Rev. 6 0-SI-SLT-081-258.1, Unit 1 Primary Water LLRT, Rev. 5 0-SI-SLT-088-259.1, Upper Personnel Airlock Overall Leak Rate Test and General Inspection,
Rev. 6
0-SI-SLT-088-259.4, Upper Personnel Airlock Interlock Operability Test, Rev. 2
1-SI-SXP-003-202.B, Motor Driven Auxiliary Feedwater Pump 1B-B Comprehensive
Performance Test, Rev. 7 1-SI-SXP-074-202.0, RHR Pump 1A-A and 1B-B Comprehensive Performance and Check Valve Test, Rev. 10
0-SI-OPS-068-137.0, Reactor Coolant System Water Inventory, Rev. 32
1-SI-OPS-088-001.0, Phase A Isolation Test, Rev. 14
1-SI-OPS-082-026A, Loss of Offsite Power with Safety Injection D/G 1A-A Test, Rev. 46 0-SI-SFT-072-138.0, Unit 1 Containment Spray - Spray Nozzle Test, Rev. 6 0-SI-SXV-063-203.2, Full Stroking of Safety Injection Cold Leg Accumulator Isolation Valves,
Rev. 1
0-SI-MIN-061-105.0, Ice Condenser Ice Weighing, Rev. 11
0-SI-SLT-067-258.2, Containment Isolation Valve Local Leak Rate Test Lower Compartment
Essential Raw Cooling Water, Rev. 13
PERs 801081, FME concern while performing air flow test during core reload
6 Attachment
Other documents
1-47W437-4, Mechanical Containment Spray System Piping, Rev. 1 1-47W437-5, Mechanical Containment Spray System Piping, Rev. 4
1-47W812-1, Flow Diagram Containment Spray System, Rev. 45
Technical Specification Surveillance Requirement 4.6.2.1.1.d and 4.6.2.1.2.b
Section 1EP2: Alert and Notification System Evaluation
Procedures and Reports
NP-REP, Appendix B, Sequoyah Nuclear Plant Radiological Emergency Plan, Rev. 101
EPFS-9, Inspection, Service, and Maintenance of the Prompt Notification System (PNS) at
Browns Ferry, Sequoyah, and Watts Bar Nuclear Plants, Rev. 8
Sequoyah FEMA REP-10 Report, Revision 2 EPDP-10, Facilitation of the ANS and Notification Tests, Rev. 6 EPDP-14, Evaluation of Changes to Alert and Notification Systems (ANS), Rev 0
Records and Data
Weekly Silent Tests, 2011-September 2013 Monthly Siren Tests, October 2011 - October 2013
Corrective Action documents
442747; During Monthly Siren Test Five Sirens Did Not Operate
521663; Siren Damaged by Storm
591666; Two ANS Sirens Failed to Operate During Monthly Test 701363; Siren Relocations Due to Land Owner Rejections 711912; Loss of DC Power Indication for ANS Siren 12
727891; Loss of DC Power Indication for ANS Siren 26
751936; Two ANS Sirens Failed to Operate During Monthly Test
Section 1EP3: Emergency Response Organization Staffing and Augmentation System
Procedures
TRN-30, Radiological Emergency Preparedness Training, Rev. 24
EPDP-3, Emergency Plan Exercises and Preparedness Drills, Rev. 7
EPDP-10, Facilitation of the Alert and Notification System and Pager Tests, Rev. 6
EPIP-3, Alert, Rev 36 EPIP-6, Activation and Operation of the Technical Support Center, Rev. 49
EPIP-7, Activation and Operation of the Operations Support Center, Rev. 28
Records and Data
SQN-EP-S-13-02, snapshot self-assessment SCBA Qualification of Site Personnel, March 2013 EPT202.000, ERO Training Plan - TSC Training, Rev. 12 EPT900.010, ERO Training Plan, ERO Fundamentals, Rev. 4
Radiological Emergency Preparedness Training Oversight Committee minutes 2012/2013
2012/2013 ERO Augmentation test results
Results of periodic ERO notification tests
Corrective Action documents
786990; TRN error in CECC qualification requirement
7 Attachment Section 1EP4: Emergency Action Level and Emergency Plan Changes
Change Packages
TVA Radiological Emergency Plan, Revs. 99 and 100
EPIP-1, "Emergency Plan Classification Matrix," Revs. 48 and 49
CECC EPIP-2, "Operations Duty Specialist Procedure for Notification of Unusual Event," Rev. 43
CECC EPIP-3, "Operations Duty Specialist Procedure for Alert," Rev. 44 CECC EPIP-4, "Operations Duty Specialist Procedure for Site Area Emergency," Rev. 45 CECC EPIP-5, "Operations Duty Specialist
Procedure for General Emergency," Rev. 50 CECC EPIP-7, "CECC Radiological Assessment Staff Procedure for Alert, Site Area
Emergency, and General Emergency," Rev. 34
TVA Radiological Emergency Plan, Rev. 101 Evacuation Time Estimate Study Update
Section 1EP5: Maintenance of Emergency Preparedness
Procedures
CECC EPIP-9, Emergency Environmental Radiological Monitoring Procedures, Rev. 49 EPDP-17, NPG Emergency Plan Effectiveness Review [10 CFR 50.54(q)], Rev. 3 NPG-SPP-7.1, On-Line Work Management, Rev. 10
NPG-SPP-18.3.5, Designated Emergency Response Equipment (DERE), Rev. 0
NPG-SPP-22.300, Corrective Action Program, Rev. 0
Records and Data
Drill and exercise reports 2011-2013 TVA Quality Assurance Audit Report SSA 1203 dated April 16, 2012
TVA Quality Assurance Audit Report SSA 1305 dated June 17, 2013
Focused Self-Assessment SQN-EP-F-13-001, NRC Inspection Preparation
SQN QA Quarterly Rating Report August 13, 2013
Corrective Action documents
571999; Maintenance Personnel Not Evacuated in a Timely Manner During REP Drill
572584; RP Was Slow to Perform Airborne Sampling During REP Drill
608785; Dose assessment error
581795; No Additional Fire Brigade Personnel Onsite During REP Drill 582858; TSC SED Filled Out Wrong Form Which Delayed CECC PAR Development 582751; MERT Failed 4 of 6 Drill Objectives
619808; RP Tech Left Team to Get Equipment During Graded Exercise
619847; Inside Van Tech Did Not Grab All Equipment Required During Graded Exercise
695758; MET Unavailable - Lessons Learned 704845; Evaluate EPIP-1 Classification of EAL 4.2 for Explosion 708940; Questioned CET Readings During Drill
711961; REP Assignment Cannot Meet 1-Hour Requirement to Respond
720352; 8 Personnel Were Not Accounted For During REP Drill
722951; KI Tablets Should Be Evaluated for Issue Earlier Under Emergency Conditions
732171; Clarify EPDP-11 regarding 10 CFR 50.54(t) requirements 751183; Wrong Pocket Ion Chambers in REP Van #3
8 Attachment
Section 2RS1: Radiological Hazard Assessment and Exposure Controls
Procedures, Guidance Documents, and Manuals
NPG-SPP-05.1.1, Alpha Radiation Monitoring Program, Rev. 3
NPG-SPP-05.6, Controlling Byproduct and Source Material, Rev. 2
NPG-SPP-22.300, Corrective Action Program, Rev. 0
RCDP-1, Conduct of Radiological Controls, Rev. 5
RCI-01, Radiation Protection Program, Rev. 78 RCI-14, Radiation Work Permit (RWP) Program, Rev. 57 RCI-15, Radiological Postings, Rev. 24
RCI-17, Control of Byproduct and Source Material, Rev. 19
RCI-18, Vacuum Cleaner Control Within the Radiologically Controlled Area, Rev. 9
RCI-21, Control of Radioactive Materials, Rev. 19 RCI-29, Control of Radiation Protection Keys, Rev. 15 RCI-101, Radiation Operations Routines, Rev. 3
RCI-106, Radiation Protection Standards and Expectations, Rev. 3
RCI-201, Radiation and Contamination Surveys, Rev. 13
RCI-202, Airborne Radioactivity Surveys, Rev. 7 RCI-204, Radiological Surveys of Equipment and Materials Leaving the RCA, Rev. 7 RCI-209, Radiological Surveys of Personnel Leaving the RCA or Protected Area, Rev. 3
RCI-301, Radionuclide Tracking and Assessment (RTA) Program, Rev. 2
RCI-412, Radiation Protection Surveys during Initial Spent Fuel Assembly Movement, Rev. 1
RCI-417, Radiological Monitoring of the Hydrogen Peroxide Injection Crud Burst, Rev. 1
RCTP-106, Special Dosimetry Operations, Rev. 2 0-TI-NUC-000-002.0, Storing Material in Spent Fuel Pool or New Fuel Vault, Rev. 21
Records and Data
Air Sample Detail Report for 10/13/2013 thru 11/5/2013, 11/5/2103
Air Sample 101713018, U1 Equipment Pit, 10/17/2013 Air Sample 101813018, U1 734' RFF GA, 10/18/2013 Air Sample 102313006, U1 Rx Head Stand, 10/23/2013
Air Sample 102313014, U1 653' 1B RHR Pump Room, 10/23/2013
Air Sample 102313023, U1 653' 1B RHR Pump Room, 10/23/2013
Air Sample 102613003, U1 Upper Rx Head O-ring Cleaning, 10/26/2013
Air Sample 110213012, U1 Upper GA, 11/2/2013 ALARA Plan 2013-010, Refueling Operations ALARA Plan 2013-011, Mechanical Maintenance Group (MMG)
ALARA Plan 2013-018, MODS - Ice Condenser/Snubbers/Insulation/Scaffolds/Painting
Instrument Calibration/Check Source Certificates: Vendor Source No. I3-328, TVA No. 2530, 7/29/2011 Vendor Source No. I3-329, TVA No. 2531, 7/29/2011 Vendor Source No. I3-330, TVA No. 2532, 7/29/2011
Vendor Source No.G4-975, TVA No. 2483, 10/9/2009
Vendor Source No. 92421, TVA No. 2571, 12/7/2012
Vendor Source No. 52736-185D2, TVA No. 2245, 5/19/2003 Instrument Calibration Records: Canberra GEM-5 Personnel Monitor, Serial No. 0909-179, 3/23/2012 and 3/18/2013 ARGOS-5AB Personnel Monitor, Instrument No. 860587, 5/11/2012 and 5/2/2013
iSolo, Instrument No. 860494, 12/6/2012 and 10/11/13
9 Attachment Small Article Monitor (Cronos 11), Instrument No. 860653, 8/17/2012 and 7/16/2013 Small Article Monitor (SAM-11), Instrument No. 860325, 7/6/2012 and 11/17/2012 List of Active SQN Temporary Shielding Request Forms (TSRFs), 11/6/2013
National Source Tracking System Annual Inventory Reconciliation Confirmation, 1/24/2013
National Source Tracking System Inventory Report, Sequoyah Nuclear Plant, 1/24/2013
RWP Dose by Work Step Report for ALARA Plans 2013-010 to 2013-021 for the period
10/14/2013 thru 11/6/2013 RWP Total Dose, Hours and Dose Rate Report for the period 10/14/2013 thru 11/5/2013 RWP Work Step Dose and Dose Rate Alarm Setpoints for RWP 13140052, 11/5/2013
RWP 13120122, U1 Seal Table work
RWP 13140002, U1 Upper Containment High Rad Area Mechanical Maintenance
RWP 13140052, HRA U1 Refueling Activities for AREVA and Boilermakers RWP 13140072, U1 HRA MODS Work: Snubbers, Scaffold, Insulation, Painting RWP 13140172, U1 Rx Head Insulation
RWP 13140252, HRA U1 Upper Containment Rx Cavity
RWP 13140352, U1 HRA Head O-Ring Surface Work & Inspection (Multibadging)
RWP 13140353, U1 Equipment Pit - LHRA Vortex Suppressors RWP 13140453, U1 Upper Containment, Rx Cavity, LHRA, CRDM duct work, (Multibadging) Radiological Survey SQN-M-20131014-23 and SQN-M-20131104-2, U1 Containment
Equipment Pit
Radiological Survey SQN-M-20131021-3, SQN-M-20131014-6, SQN-M-20131014-15, and
SQN-M-20131014-22, U1 Containment Accumulator Rooms #1, #2, #3, and #4
Radiological Survey SQN-M-20131014-32 and SQN-M-20131017-9, U1 Containment Top of Pressurizer Radiological Survey SQN-M-20130909-1 and SQN-M-20131014-8, U1 Containment Raceway
Radiological Survey SQN-M-20131014-14, SQN-M-20131014-7, and SQN-M-20131014-5, U1
Containment Steam Generator Primary Platform #1, #2, and #3
Radiological Survey SQN-M-20131014-17 and SQN-M-20131020-7, U1 Containment Inside
Polar Crane Wall Radiological Survey SQN-M-20131021-21, SQN-M-20131014-10, and SQN-M-20131014-18,
U1 Containment RCP Platform #1, #2, and #3
Radiological Survey SQN-M-20131020-16, SQN ISFSI Pad
Radiological Survey SQN-M-20121212-8, U2 Letdown Heat Exchanger Room
Radiological Survey SQN-M-20131014-26, U1 Letdown Heat Exchanger Room Radiological Survey SQN-M-20130502-11 and SQN-M-20131005-1, U1 651' Waste Evaporator Feed Pump Room
Radiological Survey SQN-M-20131018-1 and SQN-M-20131025-1, Radiochemistry Lab
Radiological Survey SQN-M-20130617-3, SQN-M-20131007-1, and SQN-M-20131104-4, Equipment Decon Room Radiological Survey SQN-M-20130823-3, and SQN-M-20131016-24, Spent Fuel Heat
Exchanger Room Radiological Survey SQN-M-20130920-5, SQN-M-20131020-4, and SQN-M-20131028-7, Spent
Fuel Pool Area Radiological Survey SQN-M-20131015-8, SQN-M-20131020-9, and SQN-M-20131024-7, 1A RHR Pump Room Radiological Survey SQN-M-20131015-11, SQN-M-20131022-8, and SQN-M-20131023-10, 1B RHR Pump Room Radiological Survey SQN-M-20131101-10, 2B RHR Pump Room
10 Attachment Radioactive Sealed Source Leak Test Certification, Source ID 0413-00-00, 7/23/09 and 1/25/10 Sequoyah Nuclear Plant Annual Radionuclide Trending and Assessment Report for 2011, 6/30/2012 Sequoyah Nuclear Plant Annual Radionuclide Trending and Assessment Report for 2012, 4/18/2013 U1R19 Radiation Protection Status Report, 11/5/2013
U1R19 RCS Shutdown Co-58 Activity Graphs (Crud Burst Cleanup), 11/5/2013 U1R19 Crud Burst Cleanup Dose Rate Trending Graphs (1A and 1B RHR Pump and Heat Exchanger rooms, and 690' and 669' Pipe Chases near RHR Lines), 11/5/2013 Waste Stream Report (10 CFR Part 61 Waste Characterization), DAW 10-22-2010, 5/11/2011
Waste Stream Report (10 CFR Part 61 Waste Characterization), DAW 3-22-2012, 11/4/2012
WO 114067330, 0-SI-RCI-000-056.0 Byproduct Material Inventory and Sealed Source Leak Test, 7/8/2013
WO 114139751, 0-SI-RCI-000-056.0 Byproduct Material Inventory and Sealed Source Leak Test, 12/17/2012
CAP Documents
Apparent Cause Evaluation PER Report, SQN PER 782859, 10/20/2013 Site Audit Report SSA1309, Radiation Protection Sequoyah Nuclear Plant, 9/16/2013
TVA Nuclear Power Group Benchmarking Report SQN-RP-I-13-BM09, 8/23/2013
PERs PER 626962 PER 629341 PER 657724
PER 659369
PER 782859
PER 788604 PER 790597 PER 793236
PER 793935
PER 799256
PER 802329
Section 2RS8: Radioactive Solid Waste Processing and Radioactive Material Handling, Storage, and Transportation
Procedures, Manuals, and Guides
Energy Solutions Cask Book for Model 8-120B USA/9168/B(U)
NPG-SPP-05.7, Radwaste Management, Rev. 0
Process Control Program (PCP), Rev. 4 Radioactive Material Shipment Manual (RMSM, Vol.II -Radioactive Material Shipment, Rev. 42 Radioactive Material Shipment Manual (RMSM, Vol.III -Radwaste Shipment, Rev. 39 RCI-06, Receipt of Radioactive Materials, Rev. 19
RCI-21 Control of Radioactive Materials, Rev. 19
RHSI-1, Packaging Dry Active Waste for Shipment to a Waste Processor/Broker or a Commercial Radwaste Burial Facility, Rev. 10 RHSI-1.1, Packaging Filters and Items of High Levels of Radiation, Rev. 6
11 Attachment RHSI-6, Bead Resin Activated Carbon Dewatering Procedure for Energy Solutions14-215 or
Smaller Liners, Rev. 8 RHSI-7, Utilization of Polyethylene High Integrity Containers (HICs) and HIC Overpacks, Rev. 9
RHSI-11, Control of Radioactive Material and Training, Rev. 6
RHSI-13, Administration and Control of Onsite Storage of Low Level Radioactive Waste, Rev. 4
RWTP-100 Attachment A, Radwaste Training Program, Rev. 3
RWTP-100, Radioactive Material/Waste Shipments, Rev. 7 RWTP-101, 10 CFR 61 Waste Characterization, Rev. 2
RWTP-102, Use of Casks, Rev. 2
0-SO-77-29, Waste Processing, Rev. 9
0-VI-RCI-077-001.0, Operating Procedure for Duratek Modular Fluidized Transfer Demineralizer System (MFTDS), Rev. 2
Shipping Records and Radwaste Data
Two Design Change Notices were reviewed and both have been accomplished. The first
moved Radwaste liquid processing from the railroad bay into the drumming room that was in
effect at the start of the period which included back to November 2010 and the second established a lift system to be used for the steam generator replacement in 2012 with a closure date of 8/13/2013.
The licensee provided several drawings delineating abandoned equipment. The inspector
chose the abandoned boric acid evaporator system to review.
Shipments: SNP-12-0111 (LQ) SNP-13-0105 (SCO)
SNP-13-0109 (Type B)
SNP-13-0307 (LSA)
SNP-13-0504 (Type A)
CAP Documents
Site Audit Report SSA1309, Radiation Protection, August 19 through August 30, 2013
Snapshot Self-Assessment Report SQN-RP-S-13-004, Radioactive Solid Waste Processing and Radioactive Material Handling, Storage and Transportation, July 29 through August 9, 2013
PERs 412285
431332
488136
635127 735591 765281
767526
783784
12 Attachment Section 4OA1: Performance Indicator Verification
Procedures, Manuals, and Guides
NSDP-29, Tracking and Trending and NRC Performance Indicators, Rev. 6
NPG-SPP-02.2, "Performance Indicator Program", Rev. 5
RCI-151, Radiation Protection Functional Area Performance Indicators, Rev. 1
PERs 621990 623246
626962
653648
655642 788604 793921
794437
Section 4OA2: Problem Identification and Resolution
Procedures
NPG-SPP-03.1, Corrective Action Program, Rev. 1
Section 4OA5: Other Activities
0-GO-17, Spent Fuel/Dry Cask Operations, Rev. 5 NPG-SPP-01.2, Administration of Site Technical Procedures, Rev. 9 NFTP-100, Fuel Selection for Dry MPC Storage, Rev. 5 completed for campaign #6 10CFR 72.48 Screening/Evaluation: EDC E22443C
SQN-DCS-300.11, Supplemental Cooling System Operation, Rev. 9
CTP-DCS-100.0, Dry Cask Storage Campaign Guidelines, Rev. 15
SQN-DCS-200.0, Dry Cask Campaign Review Program, Rev. 4 SQN-DCS-200.2, SQN-MPC-Loading and Transport Operations, Rev. 35
Attachment LIST OF ACRONYMS
ABGTS auxiliary building gas treatment system ALARA as low as reasonably achievable
ASME American Society of Mechanical Engineers
BACC boric acid corrosion control
CAP corrective action program
CCP centrifugal charging pump CDE cause determination evaluation CFR Code of Federal Regulations
CIV containment isolation valve
DAW dry active waste
DOT Department of Transportation ECCS emergency core cooling system ED electronic dosimeter
ERCW essential raw cooling water
FCV flow control valve
FME foreign material exclusion HRA high radiation areas IMC inspection manual chapter
IP inspection procedure
ISFSI independent spent fuel storage installation
ISI in-service inspection
MCC motor control center MPC multi-purpose canister NCV non-cited violation
NDE non-destructive examination
NEI Nuclear Energy Institute
PER problem evaluation report PORV power operated relief valve Radwaste radioactive waste
RCA radiologically controlled area
Rev revision
RS radiation safety RTP rated thermal power RWP radiation work permit
RWST refueling water storage tank
SDP significance determination process
SI safety injection SR service request SSC structure, system, or component
TS technical specification
TVA Tennessee Valley Authority
URI unresolved item
UT ultrasonic testing UFSAR Updated Final Safety Analysis Report WO work order