ML111170469
| ML111170469 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 04/27/2011 |
| From: | Rich D NRC/RGN-II/DRP/RPB3 |
| To: | Franke J Progress Energy Florida |
| References | |
| IR-11-002 | |
| Download: ML111170469 (31) | |
See also: IR 05000302/2011002
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
245 PEACHTREE CENTER AVENUE NE, SUITE 1200
ATLANTA, GEORGIA 30303-1257
April 27, 2011
Mr. Jon A. Franke, Vice President
Crystal River Nuclear Plant (NA1B)
15760 West Power Line Street
Crystal River, FL 34428-6708
SUBJECT: CRYSTAL RIVER UNIT 3 - NRC INTEGRATED INSPECTION REPORT
Dear Mr. Franke:
On March 31, 2011, the US Nuclear Regulatory Commission (NRC) completed an inspection at
your Crystal River Unit 3. The enclosed integrated inspection report documents the inspection
findings which were discussed on April 11, 2011, with you and other members of your staff.
The inspection examined activities conducted under your license as they related to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
Based on the results of this inspection, one self-revealing finding of very low safety significance
(Green) was identified. The finding did not involve a violation of NRC requirements.
Additionally, a licensee-identified violation, which was determined to be of very low safety
significance, is listed in this report. However, because of the very low safety significance and
because it is entered into your corrective action program, the NRC is treating this finding as a
non-cited violation (NCV) consistent with the NRC Enforcement Policy. If you contest any NCV
you should provide a response within 30 days of the date of this inspection report, with the basis
for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk,
Washington DC 20555-001; with copies to the Regional Administrator Region II; the Director,
Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-
0001; and the NRC Senior Resident Inspector at the Crystal River Unit 3 site.
In addition, if you disagree with the characterization of any finding in this report, you should
provide a response within 30 days of the date of this inspection report, with the basis for your
disagreement, to the Regional Administrator, RII, and the NRC Senior Resident Inspector at
Crystal River Unit 3.
FPC 2
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its
enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records (PARS) component of the NRCs document
system (ADAMS). Adams is accessible from the NRC Web site at http://www.nrc.gov/reading-
rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Daniel W. Rich, Chief
Reactor Projects Branch 3
Division of Reactor Projects
Docket No. 50-302
License No. DPR-72
Enclosure: Inspection Report 05000302/2011002
w/Attachment: Supplemental Information
_ML111170469________________ X SUNSI REVIEW COMPLETE
OFFICE RII:DRP RII:DRP RII:DRP RII:DRP RII:DRP RII:DRS RII:DRS
SIGNATURE SON SLM4 TXM1 by email RJR1 by email NRS2 by LFL RPC1 by email
NAME SNinh SMendez-Gonzalez TMorrissey RReyes il
NChilds LLake RCarrion
DATE 04/22/2011 04/19/2011 04/20/2011 04/20/2011 04/20/2011 04/22/2011 04/20/2011
E-MAIL COPY? YES NO YES NO YES NO YES NO YES N YES NO YES NO
OFFICE RII:DRS RII:DRP RII:DRP RII:DRP
SIGNATURE MAB7 /RA/ WGR1 by email JRS6 by email
NAME MBates DRich WRogers JSowa
DATE 04/29/2011 04/27/2011 04/20/2011 04/20/2011 04/ /2011 04/ /2011 04/ /2011
E-MAIL COPY? YES NO YES NO YES NO YES NO YES N YES NO YES NO
FPC 3
cc w/encl: Daniel R. Westcott, Supervisor
Kelvin Henderson, General Manager Licensing & Regulatory Programs
Nuclear Fleet Operations Crystal River Nuclear Plant (NA1B)
Progress Energy Electronic Mail Distribution
Electronic Mail Distribution
Joseph W. Donahue, Vice President
Brian C. McCabe Nuclear Oversight
Manager, Nuclear Oversight Progress Energy
Shearon Harris Nuclear Power Plant Electronic Mail Distribution
Progress Energy
Electronic Mail Distribution Jack E. Huegel,
Manager, Nuclear Oversight
James W. Holt, Plant General Manager Crystal River Nuclear Plant
Crystal River Nuclear Plant (NA2C) Electronic Mail Distribution
Electronic Mail Distribution
David T. Conley, Senior Counsel
Stephen J. Cahill Legal Department
Director - Engineering Nuclear Progress Energy
Crystal River Nuclear Plant (NA2C) Electronic Mail Distribution
Electronic Mail Distribution
Mark Rigsby
R. Alexander Glenn, General Counsel Manager, Support Services - Nuclear
Progress Energy Crystal River Nuclear Plant (NA2C)
Electronic Mail Distribution Electronic Mail Distribution
Jeffrey R. Swartz Senior Resident Inspector
Director Site Operations U.S. Nuclear Regulatory Commission
Crystal River Nuclear Plant Crystal River Nuclear Generating Plant
Electronic Mail Distribution U.S. NRC
6745 N Tallahassee Rd
Donna B. Alexander Crystal River, FL 34428
Manager, Nuclear Regulatory Affairs
(interim) Attorney General
Progress Energy Department of Legal Affairs
Electronic Mail Distribution The Capitol PL-01
Tallahassee, FL 32399-1050
Thomas Sapporito, Consulting Associate
(Public Correspondence Only) Bryan Koon, Director
Post Office Box 8413 Florida Division of Emergency Management
Jupiter, FL 33468 Electronic Mail Distribution
William A. Passetti, Chief Chairman
Florida Bureau of Radiation Control Board of County Commissioners
Department of Health Citrus County
Electronic Mail Distribution 110 N. Apopka Avenue
Inverness, FL 36250
FPC 4
Letter to Jon A. Franke from Daniel W. Rich dated April 27, 2011.
SUBJECT: CRYSTAL RIVER UNIT 3 - NRC INTEGRATED INSPECTION REPORT
Distribution w/encl:
C. Evans, RII
L. Douglas, RII
OE Mail
RIDSNRRDIRS
PUBLIC
RidsNrrPMCrystal River Resource
U.S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket No.: 50-302
License No.: DPR-72
Report No.: 05000302/2011002
Licensee: Progress Energy (Florida Power Corporation)
Facility: Crystal River Unit 3
Location: Crystal River, FL
Dates: January 1, 2011 - March 31, 2011
Inspectors: T. Morrissey, Senior Resident Inspector
R. Reyes, Resident Inspector
N. Childs, Resident Inspector
J. Sowa, Resident Inspector
L. Lake, Senior Reactor Inspector (Section 4OA5)
R. Carrion, Senior Reactor Inspector (Section 4OA5)
M. Bates, Senior Operations Engineer (Section R11)
Approved by: D. Rich, Chief,
Reactor Projects Branch 3
Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR 05000302/2011002; 01/01/2011-03/31/2011; Crystal River Unit 3; Licensed Operator
Requalification Program; Follow-up of Events and Notices of Enforcement Discretion
The report covered a three month period of inspection by resident inspectors, two regional
senior reactor inspectors and one regional senior operations engineer. One Green self-
revealing finding was identified. The significance of most findings is identified by their color
(Green, White, Yellow, Red) using IMC 0609, Significance Determination Process (SDP).
Findings for which the SDP does not apply may be Green or be assigned a severity level after
NRC management review. The NRCs program for overseeing the safe operation of
commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process,
Revision 4, dated December 2006.
A. NRC Identified & Self-Revealing Findings
Cornerstone: Mitigating Systems
Green: A self-revealing Green finding, associated with operating crew performance on
the simulator during facility-administered requalification examination was identified. Two
of the eight crews evaluated failed to pass their simulator examinations. As immediate
corrective action, the failed operating crews were remediated (i.e., the operating crews
were re-trained and successfully retested) prior to returning to shift. The licensee has
entered this issue into the corrective action program as Nuclear Condition Report (NRC)
450196.
The inspectors determined that the crew failures constituted a performance deficiency
based on the fact that licensed operators are expected to operate the plant with
acceptable standards of knowledge and abilities demonstrated through periodic testing
as required by 10 CFR 55.59(a)(2). Two out of eight crews of licensed operators failed
to demonstrate a satisfactory understanding of the required actions and mitigating
strategies required to safely operate the facility under normal, abnormal, and emergency
conditions. The finding is greater than minor because the performance deficiency
potentially affects the Human Performance attribute of the Mitigating Systems
cornerstone objective to ensure the availability, reliability, and capability of systems that
respond to initiating events to prevent undesirable consequences. Specifically, the
finding reflected the crews potential inability to take timely actions in response to actual
abnormal and emergency conditions. The cause of this finding was directly related to
the cross-cutting aspect of personnel training and qualifications in the Resources
component of the Human Performance area, in that the licensee failed to ensure the
adequacy of the training provided to operators to assure nuclear safety. (H.2(b))
(Section 1R11)
B. Licensee Identified Violations
One violation of very low safety significance, which was identified by the licensee, has
been reviewed by the inspectors. Corrective actions taken or planned by the licensee
have been entered into the licensees corrective action program. The violation and
corrective action tracking number is listed in Section 4OA7 of this report.
Enclosure
REPORT DETAILS
Summary of Plant Status:
Crystal River Unit 3 began the inspection period in Mode 5 (< 200oF). Unit 3 remained in Mode
5 for the remainder of the inspection period.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
1R01 Adverse Weather Protection
.1 Adverse Weather Protection: Tornado Watch / Warning
a. Inspection Scope
On January 25, 2011, and again on March 30-31 the inspectors evaluated the licensees
preparations when the site was informed of being in a tornado watch then subsequently
in a tornado warning. The licensee implemented emergency management procedure
EM-220, Violent Weather, for the tornado watches and warnings. The inspectors walked
down the outside protective area to ensure actions required by EM-220 were
implemented. This constituted two samples representing observation of adverse
weather protection activities.
b. Findings
No findings were identified. The tornado watches and warnings expired with no violent
weather or tornado formation near the site.
1R04 Equipment Alignment
.1 Partial Equipment Walkdowns
a. Inspection Scope
The inspectors performed walkdowns of the critical portions of the selected trains to
verify correct system alignment. The inspectors reviewed plant documents to determine
the correct system and power alignments, and the required positions of select valves
and breakers. The inspectors verified that the licensee had properly identified and
resolved equipment alignment problems that could cause initiating events or impact
mitigating system availability. The inspectors verified the following two partial system
alignments in system walkdowns using the listed documents:
Enclosure
4
- A train nuclear service water (SW) and A train raw water (RW) systems, using
operating procedure OP-408, Nuclear Services Cooling System, while B trains of SW
and RW were out of service for scheduled maintenance
- Emergency diesel generator EGDG-1B using OP-707, Operation of the Emergency
Diesel Generators, while the EGDG-1A was out of service for surveillance testing
b. Findings
No findings were identified.
1R05 Fire Protection
.1 Fire Area Walkdowns
a. Inspection Scope
The inspectors walked down accessible portions of the plant to assess the licensees
implementation of the fire protection program. The inspectors checked that the areas
were free of transient combustible material and other ignition sources. Also, fire
detection and suppression capabilities, fire barriers, and compensatory measures for fire
protection problems were verified. The inspectors checked fire suppression and
detection equipment to determine whether conditions or deficiencies existed which could
impair the function of the equipment. The inspectors selected the areas based on a
review of the licensees probabilistic risk assessment. The inspectors also reviewed the
licensees fire protection program to verify the requirements of Final Safety Analysis
Report (FSAR) Section 9.8, Plant Fire Protection Program, were met. Documents
reviewed are listed in the attachment. The inspectors toured the following five areas
important to safety:
- Remote shut down panel, and the A and B emergency service 4160-Volt switch gear
rooms
- Sea water room 95 elevation auxiliary building
- Intermediate building 95 elevation emergency feed water pump EFP-1 and EFP-2
area
- Unit 3 main control room
- Control complex control rod relay cabinet room
b. Findings
No findings were identified.
Enclosure
5
.2 Annual Fire Drill
a. Inspection Scope
On January 18 and on January 23, the inspectors observed two separate licensee fire
brigade responses to a simulated fire. Both drills involved a fire in the turbine building
480V unit switchgear room 95 elevation. The inspectors checked the brigades
communications, ability to set up and execute fire operations, and their use of fire-
fighting equipment. The inspectors verified compensatory actions were in place to
ensure that additional alarms which may be received during the drill were addressed.
Additionally, the inspectors verified that the licensee considered the aspects as
described below when the brigade conducted the firefighting activities and during the
post drill critique. The inspectors attended the post-drill critiques to check that the
licensees drill acceptance criteria were met and that any discrepancies were discussed
and resolved. Administrative instruction AI-2205, Administration of CR-3 Fire Brigade,
was reviewed to assure that acceptance criteria were evaluated and deficiencies were
documented and corrected. In addition, the inspectors reviewed the storage, training,
expectations for use and maintenance associated with the self-contained breathing
apparatus (SCBA) program. This inspection completed one sample representing
observation of selected fire drills. Documents reviewed are listed in the attachment.
The inspectors observed that:
- The brigade, including the fire team leader, had a minimum of five members.
- Members set out designated protective clothing and properly donned gear.
- SCBA were available and properly used.
- Control room personal verified fire location, dispatched fire brigade and sounded
alarms. Emergency action levels were declared and notifications were completed.
- Fire brigade leader as well as the control room senior reactor operator had copies of
the pre-fire plans.
- Brigade leader maintained control: Members were briefed, discussed plan of attack,
received individual assignments, and completed communications checks. Plan of
attack discussions were consistent with pre-fire plans.
- Fire brigade arrived at the fire scene in a timely manner, taking the appropriate
access route specified in the strategies and procedures.
- Control and command was set up near the fire scene and communications were
established with the control room and the fire brigade members.
- Effectiveness of radio communication between the command post, control room,
plant operators and fire brigade members.
- Fire hose lines reached all necessary fire hazard locations, were laid out without flow
constrictions, and were simulated as being charged with water.
- The fire area was entered in a controlled manner following the two person rule.
- The fire brigade brought sufficient fire-fighting equipment to the scene to properly
perform its fire-fighting duties.
- The fire brigade checked for fire victims and fire propagation into other areas.
- Effective smoke removal operations were simulated in accordance with the pre-fire
plan.
- The fire-fighting plan strategies were utilized.
- The drill scenario was followed, and the drill acceptance criteria were met.
- All firefighting equipment was returned to a condition of readiness.
Enclosure
6
b. Findings
No findings were identified.
1R06 Flood Protection Measures
Internal Flood Protection
a. Inspection Scope
The Inspectors reviewed the Crystal River Unit 3, FSAR, Chapter 2.4.2.4, Facilities
Required for Flood Protection, and the Crystal River Unit 3 design basis documents that
depicted protection for areas containing safety-related equipment to identify areas that
may be affected by internal flooding. A walkdown of the emergency feed pump EFP-1
and EFP-2 area was conducted to ensure that flood protection measures were in
accordance with design specifications. Specific plant attributes that were checked
included structural integrity, sealing of penetrations, and operability of sump systems.
b. Findings
No findings were identified
1R07 Heat Sink Performance
Annual Review
a. Inspection Scope
The inspectors observed maintenance personnel perform heat exchanger inspections
and cleaning for the service water heat exchanger SWHE-1B. The inspector reviewed
the as-found conditions when the heat exchanger was opened for inspection and tube
cleaning to verify the heat exchanger was in an acceptable condition to perform its
design function. In addition, the inspectors observed heat exchanger maintenance that
included tube replacement and recoating of the end bell and channel head. The
documents reviewed are listed in the attachment.
b. Findings
No findings were identified.
1R11 Licensed Operator Requalification Program
.1 Resident Inspector Quarterly Review
a. Inspection Scope
On February 1, the inspectors observed and assessed licensed operator crew response
and actions for the Crystal River Unit 3 licensed operator simulator evaluated session
SES-161. Session SES-161 involved two major transients: B train steam generator tube
Enclosure
7
leak; and a spurious reactor trip. The plant conditions degraded to a point where the
licensee entered an Alert emergency classification. The inspectors observed the
operators use of abnormal procedures AP-545, Plant Runback; and AP-510 Rapid
Power Reduction. Additionally, emergency operating procedures used during the
scenario included EOP-02, Vital System Status Verification and EOP-06, Steam
Generator Tube Rupture. The operators actions were verified to be in accordance with
the above procedures. Event classification and notifications were verified to be in
accordance with emergency management procedure EM-202, Duties of the Emergency
Coordinator. The simulator instrumentation and controls were verified to closely parallel
those in the actual control room. The inspectors attended the management crew critique
and evaluation to verify the licensee had entered any adverse conditions into the
corrective action program. The inspectors evaluated the following attributes related to
crew performance:
- Clarity and formality of communication
- Ability to take timely action to safely control the unit
- Prioritization, interpretation, and verification of alarms
- Correct use and implementation of abnormal and emergency operation procedures
and emergency plan implementing procedures
- Control board operation and manipulation, including high-risk operator actions
- Oversight and direction provided by supervision, including ability to identify and
implement appropriate technical specification actions, regulatory reporting
requirements, and emergency plan classification and notification
- Crew overall performance and interactions
b. Findings
No findings were identified.
.2 Biennial Review by Regional Inspector
a. Inspection Scope
The inspector reviewed the facility operating history and associated documents in
preparation for this inspection. During the week of February 21, 2011, the inspector
reviewed documentation, interviewed licensee personnel, and observed the
administration of operating tests associated with the licensees operator requalification
program. Each of the activities performed by the inspector was done to assess the
effectiveness of the facility licensee in implementing requalification requirements
identified in 10 CFR Part 55, Operators Licenses. Evaluations were also performed to
determine if the licensee effectively implemented operator requalification guidelines
established in NUREG-1021, Operator Licensing Examination Standards for Power
Reactors, and Inspection Procedure 71111.11, Licensed Operator Requalification
Program. The inspector also evaluated the licensees simulation facility for adequacy
for use in operator licensing examinations using ANSI/ANS-3.5-1998, American
National Standard for Nuclear Power Plant Simulators for Use in Operator Training and
Examination. The inspector observed a crew during the performance of the operating
tests. Documentation reviewed included written examinations, Job Performance
Measures (JPMs), simulator scenarios, licensee procedures, on-shift records, simulator
modification request records, simulator performance test records, operator feedback
Enclosure
8
records, licensed operator qualification records, remediation plans, and medical records.
The records were inspected using the criteria listed in Inspection Procedure 71111.11.
Documents reviewed during the inspection are listed in the Attachment.
On February 25, 2011, the licensee completed the comprehensive biennial
requalification written examinations and annual requalification operating tests required to
be administered to all licensed operators in accordance with 10 CFR 55.59(a)(2). The
inspector performed an in-office review of the overall pass/fail results of the written
examinations, individual operating tests and the crew simulator operating tests. These
results were compared to the thresholds established in Manual Chapter 0609 Appendix
I, Operator Requalification Human Performance Significance Determination Process.
b. Findings
Introduction: A self-revealing Green finding, associated with operating crew
performance on the simulator during facility-administered requalification examination
was identified when two of eight crews failed the simulator portion of the facility-
administered annual operating test. Based on the licensees successful remediation and
subsequent re-testing of individuals who failed the simulator portion of the annual
operating test, no violation of regulatory requirements occurred.
Description: During the facility-administered annual operating test of licensed operators,
covering the period from January 19, to February 24, 2011, the licensees training staff
evaluated crew performance during dynamic scenarios. The evaluations were
performed using TRN-NGGC-0420, Conduct of Simulator Training and Evaluation,
Rev. 0. Facility results of crew performance indicated that two of eight crews (25
percent) did not pass their simulator exam. The licensees training staff determined that
two crews failed to meet the criteria for satisfactory performance of critical tasks. The
crew failures of simulator operational evaluations on the 2011 annual operating test have
been addressed in the licensees corrective action program with nuclear condition report
(NCR) 450196.
Analysis: The inspector determined that the crew failures constituted a performance
deficiency based on the fact that licensed operators are expected to operate the plant
with acceptable standards of knowledge and abilities demonstrated through periodic
testing as required by 10 CFR 55.59(a)(2). Two out of eight crews of licensed operators
failed to demonstrate a satisfactory understanding of the required actions and mitigating
strategies required to safely operate the facility under normal, abnormal and emergency
conditions. The finding was greater than minor because the performance deficiency was
associated with the Human Performance attribute of the Mitigating Systems cornerstone
and affected the cornerstone objective of ensuring the availability, reliability, and
capability of systems that respond to initiating events to prevent undesirable
consequences. Specifically, the finding reflected the crews potential inability to take
timely actions in response to actual abnormal and emergency conditions.
The perceived risk associated with the number of crews failing the annual operating test
was provided in the Simulator Operational Evaluation matrix of NRC Manual Chapter 0609, Appendix I, Licensed Operator Requalification Significance Determination
Process (SDP). The matrix was entered based on the number of crews that took the
simulator test (eight) and the number of crews with unsatisfactory performance (two).
Based on a crew failure rate of 25 percent on the simulator portion of the annual
Enclosure
9
operating test, the fact that the failed operating crews were remediated (i.e., the
operating crews were re-trained and successfully re-tested) prior to returning to shift,
and because there was no similar finding the previous year, this finding was
characterized by the SDP as having a very low safety significance, or Green. The cause
of this finding was directly related to the cross-cutting aspect of personnel training and
qualifications in the Resources component of the Human Performance area, in that the
licensee failed to ensure the adequacy of the training provided to operators to assure
nuclear safety. (H.2(b))
Enforcement: This finding does not involve enforcement action because no regulatory
requirement violation was identified. Because this finding does not involve a violation
and has a very low safety significance, it is identified as FIN 5000261/2011002-01, Two
of Eight Operating Crew Failures on the Simulator Operational Evaluation Portion of the
2011 Annual Requalification Operating Test.
1R12 Maintenance Effectiveness
a. Inspection Scope
The inspectors reviewed the licensees effectiveness in performing routine maintenance
activities. The review included the identification, scope, and handling of degraded
equipment conditions, as well as common cause failure evaluations, and the resolution,
of historical equipment problems. For those systems, structures, and components within
the scope of the Maintenance Rule (MR) per 10 CFR 50.65 (a)(1) and (a)(2),
classifications were justified in light of the reviewed degraded equipment condition. The
documents reviewed are listed in the attachment. The inspectors conducted this
inspection for the following four items:
- System Engineering (SE) report SE11-0006, Remove Spent Fuel Pump Motor
Cooling system (AH-XG) from the Scope of the Maintenance Rule
- NCR 434362, Raw water pump RWP-2B reduced seal flush flow
- SE11-0014, Nuclear Instrumentation Source Range to Return to (a)(2)
- Inspector review of licensees preventative maintenance program associated with
components that have a recommended vendor service life. This completes the NRC
review utilizing Operating Experience Smart Sample (OpESS) FY 2010-01 Recent
Inspection Experience for Components Installed Beyond Vendor Recommended
Service Life.
b. Findings
No findings were identified.
1R15 Operability Evaluations
a. Inspection Scope
The inspectors reviewed the following two NCRs to verify operability of systems
important to safety was properly established, that the affected components or systems
remained capable of performing their intended safety function, and that no unrecognized
increase in plant or public risk occurred. The inspectors determined if operability of
Enclosure
10
systems or components important to safety was consistent with Improved Technical
Specifications (ITS), the FSAR, 10 CFR Part 50 requirements, and when applicable,
NRC Inspection Manual, part 9900, Technical Guidance, Operability Determinations &
Functionality Assessments for Resolution of Degraded or Nonconforming Conditions
Adverse to Quality or Safety. The inspectors reviewed licensee NCRs, work schedules,
and engineering documents to check if operability issues were being identified at an
appropriate threshold and documented in the corrective action program, consistent with
10 CFR 50, Appendix B requirements and licensee procedure CAP-NGGC- 200,
Condition Identification And Screening Process. The documents reviewed are listed in
the attachment.
- NCR 447026, SWHE-1B front and back channel heads corroded with several areas
below the vendor recommended limit
- NCR 436065, Evaluate makeup system valve MUV-163 for possible degradation
b. Findings
No findings were identified.
1R18 Plant Modifications
Permanent Plant Modifications
a. Inspection Scope
The inspectors reviewed the engineering change (EC) 76007, Emergency Feedwater
Initiation and Control Flow Circuit, to verify it met the requirements of engineering
procedures EGR-NGGC-0003, Design Review Requirements, and EGR-NGGC-0005,
Engineering Change. The inspectors observed the as-built configuration of the
modification and observed installation, and observed testing activities associated with
the modification. Documents reviewed included surveillance procedures, design and
implementation packages, work orders (WOs), system drawings, corrective action
documents, applicable sections of the FSAR, ITS, and design basis information. Post
maintenance testing data and acceptance criteria were reviewed. The inspectors
verified that issues found during the course of the installation and testing associated with
the modification were entered and properly dispositioned in the licensees corrective
action program.
b. Findings
No findings were identified.
1R19 Post Maintenance Testing
a. Inspection Scope
The inspectors either observed or reviewed post-maintenance test results as
appropriate, for selected risk significant systems to verify whether: (1) testing was
adequate for the maintenance performed; (2) acceptance criteria were clear, and
adequately demonstrated operational readiness consistent with design and licensing
Enclosure
11
basis documents; (3) test instrumentation had current calibrations, range, and accuracy
consistent with the application; (4) tests were performed as written with applicable
prerequisites satisfied, and (5) equipment was returned to the status required to perform
its safety function. The five post-maintenance tests reviewed are listed below:
- Surveillance procedure SP-354B, Monthly Functional Test of the Emergency Diesel
Generator EGDG-1B, after performing maintenance per work orders (WOs)
01706861 and 01435520
- SP-344B, RWP-2B, SWP-1B and Valve Surveillance, after performing maintenance
on RWP-2B and SWP-1B per WOs 1852978, 1332475, 1332477, 1848527 and
1852978
- Performance Test PT-445, Control Rod Programming Verification, after performing
maintenance per WO 1893481
performing maintenance per WO 1691112
- SP-344A, RWP-2A, SWP-1A and Valve Surveillance, after performing maintenance
per WO 1063292
b. Findings
No findings were identified.
1R20 Refueling and Outage Activities
Steam Generator Replacement Refueling Outage (RFO16)
a. Inspection Scope
On September 26, 2009, the unit was shutdown for a planned steam generator
replacement refueling outage. The previous quarters NRC inspection activities in this
area were documented in NRC integrated inspection report 05000302/2010005. During
this quarter, the inspectors observed and monitored licensee controls over the refueling
outage activities listed below. Documents reviewed are listed in the Attachment.
- Outage related risk assessment monitoring
- Controls associated with shutdown cooling, reactivity management, electrical power
alignments, containment closure, and spent fuel pool cooling
- Implementation of equipment clearance activities
b. Findings
No findings were identified
1R22 Surveillance Testing
a. Inspection Scope
The inspectors observed surveillance tests or reviewed the test results for the six
surveillance tests listed below to verify that ITS surveillance requirements were followed
and that test acceptance criteria were properly specified. The inspectors verified that
Enclosure
12
proper test conditions were established as specified in the procedures, that no
equipment preconditioning activities occurred, and that acceptance criteria had been
met. Additionally, the inspectors verified that equipment was properly returned to
service and that proper testing was specified and conducted to ensure that the
equipment could perform its intended safety function.
In-Service Test:
- SP-340A, RWP-3A, DCP-1A and Valve Surveillance
Surveillance Test:
- PT-315, Remote Shutdown Relay Operability
- SP-524, Battery Modified Performance Discharge Test (A train only)
- SP-108, Reactor Trip Module And Control Rod Drive Trip Functional Test
(sections 4.3 and 4.4 only)
and 4.5)
b. Findings
No findings were identified.
1EP6 Drill Evaluation
a. Inspection Scope
The inspectors observed two emergency response activities to verify the licensee was
properly classifying emergency events, making the required notifications, and
appropriate protective action recommendations. The inspectors assessed the licensees
ability to classify emergent situations and make timely notification to State and Federal
officials in accordance with 10 CFR Part 50.72. Emergency activities were verified to be
in accordance with the Crystal River Radiological Emergency Response Plan, Section
8.0, Emergency Classification System, and 10 CFR Part 50, Appendix E. Additionally,
the inspectors verified that adequate licensee critiques were conducted in order to
identify performance weaknesses and necessary improvements.
- February 1, license operator simulator evaluated session, SES-161, involving a
steam generator tube rupture and a spurious reactor trip
- March 1, Crystal River Unit 3 2011 radiological emergency response training drill.
The drill scenario included equipment failures on the operating reactor that caused
the licensee to make emergency classifications and notifications and activate the
technical support center (TSC) and the emergency operating facility (EOF). The
inspectors observed the drill activities at the Unit 3 simulator, TSC, and the EOF.
The inspectors attended the drill critiques at the TSC and EOF to verify the licensee
had adequately identified any performance weaknesses and necessary
improvements.
Enclosure
13
b. Findings
No findings were identified.
4. OTHER ACTIVITIES
4OA2 Problem Identification and Resolution
.1 Daily Review
a. Inspection Scope
As required by Inspection Procedure 71152, Identification and Resolution of Problems,
and in order to help identify equipment failures or specific human performance issues for
follow-up, the inspectors performed a daily screening of items entered into the licensees
corrective action program (CAP). This review was accomplished by attending daily plant
status meetings, interviewing plant operators and applicable system engineers, and
accessing the licensees computerized database.
b. Findings
No findings were identified.
.2 Annual Sample Review
a. Inspection Scope
The inspectors selected several NCRs documenting NRC identified deficiencies
associated with spent fuel pool foreign material exclusion area (FMEA) controls for a
detailed review and discussion with the licensee. These deficiencies were identified by
the inspectors over the last several months. The NCRs reviewed are listed in the
attachment. The NCRs were written to address improper FMEA log entries, material in
the area not properly logged, and expansion of the FMEA without verifying proper
cleanliness of the expanded area. The inspectors verified that the issues were
completely and accurately identified in the licensees corrective action program, safety
concerns were properly classified and prioritized for resolution, the cause determination
was sufficiently thorough, and appropriate corrective actions were initiated. The
inspectors also evaluated the NCRs using the requirements of the licensees CAP as
delineated in corrective action procedure CAP-NGGC-200, Condition Identification and
Screening Process.
b. Findings and Observations
No findings were identified. The inspectors noted that the licensee took immediate and
appropriate actions to address each of the identified deficiencies. The licensees
corrective action to require an FME monitor for all entries into the spent fuel pool FMEA
should prevent recurrence of similar issues. The inspectors determined that there were
no identified consequences associated with the FMEA issues identified.
Enclosure
14
.3 Semi-Annual Trend Review
a. Inspection Scope
As required by Inspection Procedure 71152, Identification and Resolution of Problems,
the inspectors performed a review of the licensees CAP 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 considered the
results of daily inspector CAP item screening discussed in section 4OA2.1 above, plant
status reviews, plant tours, and licensee trending efforts. The inspectors review
nominally considered the six month period of October 2010 through March 2011. The
review also included issues documented in the licensees Plant Health Committee Site
Focus List - March 2011, various departmental CAP Rollup & Trend Analysis reports for
the 4th quarter 2010, various nuclear assessment section reports and maintenance rule
(MR) reports. Corrective actions associated with a sample of the issues identified in the
licensees corrective action program were reviewed for adequacy.
b. Assessment and Observations
No findings were identified. The inspectors evaluated the licensees trend methodology
and observed that the licensee had performed a detailed review. The inspectors review
of licensee performance over the last six months noted one negative trend associated
with spent fuel pool FMEA controls. The licensee is aware of the negative trend and has
implemented appropriate corrective actions. Additional detail can be found in section
4OA2.2.
4OA3 Follow-up of Events and Notices of Enforcement Discretion
(Closed) LER 05000302/2010-001-00, -01, -02 As-Found Cycle 16 Pressurizer Code
Safety Valve Setpoints Outside Improved Technical Specification Limit
With Crystal River Unit 3 in No Mode (core off loaded), the licensee determined that the
as-found lift setpoints of the two pressurizer code safety valves (PCSV) removed after
the September 2009 unit shut down were outside Improved Technical Specification (ITS)
limits. ITS 3.4.9 requires that two PCSVs shall be operable in Modes 1, 2, and 3. To be
operable, the lift setpoints must be within +/- 2 percent of 2500 psig. The lift setpoints for
the PCSVs were found to be 5.32 percent and 2.08 percent above the ITS setpoint
respectively. The licensee concluded that both PCSVs were inoperable for a period
longer than allowed by plant ITS. A root cause could not be determined.
The licensee identified a selected cause associated with the licensees failure to manage
vendor quality. The licensee failed to provide proper relief valve specifications to the
vendor including a detailed testing procedure, repair plan and acceptance criteria.
Corrective actions planned or completed include: changing the as-left setpoint to +0/-1
percent of the nominal setpoint; installing PCSVs with +0/-1 percent of nominal setpoint
prior to unit startup; creation of a test procedure for steam testing the PCSV to meet the
licensees standards; and revision of specifications associated with PCSV repairs.
The finding was evaluated under the Significance Determination Process (SDP) using
Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 - Initial Screening and
Characterization of Findings, and was determined to degrade the RCS barrier under the
Enclosure
15
Barriers Cornerstone. Utilizing Table 4a, Attachment 0609.04, the issue was screened
as needing an SDP Phase III evaluation. When notified that an SDP Phase III
evaluation was required, the licensee contracted with Areva NP Inc. to analyze the
impact of high as-found PCSV setpoints on peak reactor coolant system (RCS) pressure
for the most limiting accident transients.
Areva Technical Data Record 12-9154488-000, CR-3 Pressurizer Code Safety Valve
Analysis for Licensee Event Report, concluded that for the most limiting transients
(startup accident, loss of feed water and feed water line break), the peak RCS pressure
remained below the acceptance criteria for each transient and would not impact RCS
integrity. Revision 02 of the LER documents this Areva NP Inc. analysis. With this
additional information, the inspectors in conjunction with the Regional NRC Senior
Reactor Analyst (SRA) concluded that the PCSVs, with their as-found setpoints outside
of ITS limits, would have performed their safety function and a formal SDP Phase III
evaluation would not be required. Therefore, the finding was determined to be of very
low safety significance (Green). The inspectors determined that this violation of ITS
3.4.9, Pressurizer Safety Valves, met the criteria for a licensee-identified violation. The
enforcement aspects of the violation are discussed in Section 4OA7. This LER is closed.
b. Finding
No findings were identified.
4OA5 Other Activities
.1 Quarterly Resident Inspector Observations of Security Personnel 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 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 reviews and inspection activities.
b. Finding
No findings were identified.
Enclosure
16
.2 Steam Generator Replacement Project and Containment Wall Repair (IP 50001)
a. Inspection Scope
Steam Generator Replacement Project Activities
The inspectors reviewed the following issues:
Bulges of Liner Plate The licensee developed an engineering calculation to evaluate
bulges in the CR3 liner plate. It was directed at determining an apparent cause for the
bulges and establishing an analytically-based acceptance criterion for the bulges within
the CR3 design basis. The analyses included finite element modeling of the liner and
the associated anchorage to the concrete containment structure. The apparent cause
for the bulges was determined to be a combination of elements, including geometrical
imperfections in the original liner plate during construction. The calculations considered
worst case configurations and a threshold for bulge size was established considering the
effects that occur due to normal operation and accident conditions. The primary
variables in the bulge evaluation were determined to be bulge size and thermal loading.
The calculation found that the bulges have an insignificant effect on the response of the
structure due to various load combinations. The current bulges are bounded by the
acceptance criteria in the analysis. To ensure that conditions are acceptable in the
future, the licensee planned to include bulge surveillance in the international welding
engineer (IWE) program. The licensee also planned to validate the effects of
retensioning on bulge size by measurement and evaluation of a representative sample
before initiating Structural Integrity Test (SIT) pressurization as well as performing a
complete baseline scan after completion of the SIT.
50.59 Evaluation The inspectors reviewed the licensees evaluation of the containment
building modification resulting from the introduction of the construction opening and its
subsequent restoration with respect to requirements of 10 CFR, § 50.59, Changes,
Tests and Experiments, to determine whether the design bases, licensing bases, and
performance capability of the containment had been degraded through the modification
and to determine whether the design and license basis documentation used to support
changes reflected the design and license basis of the facility after the change had been
made. This evaluation remained ongoing pending completion of containment repairs,
completion of tendon retensioning, completion of post modification testing, and
subsequent validation of design parameters.
Vertical Cracks of Containment Building The licensee determined that the vertical
cracks discovered on the exterior wall of the Containment Building would close as the
buildings tendons were retensioned. The inspectors walked down selected vertical
cracks being monitored by the licensee to evaluate their condition. The licensee had
measured the cracks periodically and determined that they were closing as the tendon
retensioning process continued. The inspectors also visited the tendon control center
where the retensioning process was controlled and which housed the acoustic
monitoring and strain gage instrumentation and interviewed personnel in the center to
better understand the operation of the systems being used and how the information
obtained was interpreted. Inspection in this area remained ongoing pending completion
of tendon retensioning and subsequent validation of design parameters.
Enclosure
17
Tendon Retensioning Activities The inspectors reviewed the licensees retensioning
plans, procedures, and drawings. Retensioning activities began on January 4, 2011.
The inspectors observed some of the retensioning work on selected hoop tendons as it
was being performed to verify that the work was being conducted per approved
procedures.
Structural Integrity Test (SIT) / Integrated Leak Rate Test (ILRT) Preparations The
inspectors interviewed licensee personnel responsible for the planned SIT/ILRT to
determine the status of the test preparations, walked down the containment building to
verify the locations of the extensometers to be used to measure the containment
movements during the SIT/ILRT, and discussed the licensees procedures to assure that
they conformed to industry standards and ASME Code requirements.
Events of March 14, 2011 On the afternoon of March 14, 2011, the licensee had
completed the first retensioning sequence (Sequence #100, Hoop Tendons 42H41,
62H41, and 64H41) of the final pass (Pass 11). Per procedure, the licensee was waiting
for the containment building to stabilize before beginning the next sequence and was
monitoring the structural behavior of the containment building via acoustical emissions
monitors and strain gauges, specifically placed at various points of the structure to
detect any abnormal/unexpected response to tendon retensioning. During this
monitoring period, the strain gauges indicated an increase in strain and then failed high,
and the acoustic monitors indicated a high level of acoustic activity in the bay bordered
by Buttresses 5 and 6 (Bay 5-6). Sound coming from the bay was reported to sound like
popcorn popping by workers in the area. The phenomenon reportedly lasted for
approximately twenty minutes. The licensee utilized impulse response (IR) non-
destructive examination (NDE) techniques to determine the condition of the wall in Bay
5-6. The IR scans of the bay determined that there were numerous indications
consistent with a delamination. By the end of the inspection period, the licensee had
determined that the delamination was extensive in Bay 5-6 and was continuing to
evaluate the condition of the containment structure.
b. Findings
No findings were identified.
4OA6 Exit
Exit Meeting Summary
On April 11, 2011, the resident inspectors presented the inspection results to Mr. J.
Franke, Site Vice President and other members of licensee management. The
inspectors confirmed that proprietary information was not provided or examined during
the inspection.
4OA7 Licensee Identified Violations
The following issue of very low safety significance (Green) was identified by the licensee
and was a violation of NRC requirements. This issue met the criteria of the NRC
Enforcement Policy, NUREG-1600, for being dispositioned as a non-cited violation.
Enclosure
18
Improved Technical Specification (ITS) 3.4.9 states that two pressurizer code safety
valves (PCSVs) shall be operable in Modes 1, 2 and 3. To be operable, the lift setpoints
must be within +/- 2 percent of 2500 psig. Contrary to the above, on September 1, 2010
and on October 5, 2010, Progress Energy was notified that the as-found lift setpoints of
PCSVs RCV-9 and RCV-8 were outside ITS setpoint limits, respectively. The as-found
lift setpoint of RCV-9 was 5.32 percent above the lift setpoint and RCV-8 was 2.08
percent above the lift setpoint. The licensee identified a selected cause associated with
the licensees failure to manage vendor quality. The performance deficiency, failure to
provide proper relief valve specifications to the vendor, was determined to be greater
than minor because if left uncorrected, the performance deficiency would have the
potential to lead to a more significant safety concern regarding the integrity of the reactor
coolant system (RCS) barrier during plant transients. Corrective actions planned or
completed include: changing the as-left setpoint to +0/-1 percent of the nominal setpoint;
installing PCSVs with +0/-1 percent of nominal setpoint prior to unit startup; creation of a
test procedure for steam testing the PCSV to meet the licensees standards; and
revision of specifications associated with PCSV repairs. As documented in Section
4OA3, the finding was determined to be of very low safety significance (Green) because
there was no loss of safety function due to the lift setpoints being outside of the ITS limit.
This issue was documented in the licensees corrective action program as NCR 426852.
ATTACHMENT: SUPPLEMENTAL INFORMATION
Enclosure
KEY POINTS OF CONTACT
Licensee personnel:
J. Holt, Plant General Manager
D. Douglas, Manager, Maintenance
S. Cahill, Director, Engineering
J. Huegel, Manager, Nuclear Oversight
P. Dixon, Manager Training
B. Wunderly, Manager, Operations
D. Westcott, Supervisor, Licensing
B. Akins, Superintendent, Radiation Protection
C. Poliseno, Supervisor, Emergency Preparedness
R, Wiemann, Acting Director, Engineering
I. Wilson, Manager Outage and Scheduling
J. Franke, Vice President, Crystal River Nuclear Plant
M. Van Sicklen, Superintendent Operations Training
R. Llewellyn, Supervisor - Operations Continuing Training
NRC personnel:
D. Rich, Chief, Branch 3, Division of Reactor Projects
LIST OF ITEMS OPENED, CLOSED
Opened and Closed
05000302/2011002-01 FIN Operating Crew Failures on the 2011 Annual
Requalification Operating Test (Section 1R11.2)
Closed
05000302/2010001-00, -01, -02 LER As-Found Cycle 16 Pressurizer Code Safety
Valve Setpoints Outside Improved technical
Specification Limit
Attachment
LIST OF DOCUMENTS REVIEWED
Section 1R05: Fire Protection
Procedures
AI-2205A, Pre Fire Plan - Control Complex
AI-2205B, Pre Fire Plan - Turbine Building
AI-2205C, Pre Fire Plan - Auxiliary Building
Al -2205F, Pre Fire Plan - Miscellaneous buildings and Components
SP-804, Surveillance of Plant Fire Brigade Equipment
HPP-500, Respiratory Protection Program
HPP-502, Respiratory Equipment Inspection and Maintenance
AP-880, Fire Protection
TRN-NGGC-0010, Fitness-for-Duty, Plant Access, Radiation Worker, and Respiratory
Protection Training
Work Requests
WR 446419, Fan Failure
WR 458311, FH-11 hole in roof
WR 458172, FSV-304 packing leak
Work Orders
WO 1870980, Monthly SP-804, Surveillance Of Plant Fire Brigade Equipment, Dated 2/5/11
Section 1R07: Heat Sink Performance
Procedures
PM-275, General Preventative Maintenance
MP-299, Heat Exchanger Tube Plugging and Tube Removal/Replacement
Work Orders
WO 976734 SWHE-1B tube replacement
Attachment
3
Section 1R11: Licensed Operator Requalification
Records
License Reactivation Packages (3)
LORP Training Attendance records (5)
Medical Files (10)
Remedial Training Records (2)
Remedial Training Examinations (2)
Feedback Summaries (10)
Simulator Trouble Reports (11)
Written Examinations
2011 Biennial Written Exam C1 SRO
Procedures
TRN-NGGC-0420, Conduct of Simulator Training and Evaluation, Rev. 0
TRN-NGGC-0002, Performance Review and Remedial Training, Rev. 0
TAP-403, Conduct of Written Examinations, Rev. 17
TAP-413, Conduct of Operator Continuing Training, Rev. 14
TAP-428, Simulator Scenario-Based Testing, Rev. 1
TPP-206, Simulator Program, Rev. 9
TPP-422, Simulator Maintenance, Rev. 4
Simulator Tests
Validation of Simulator to Plant ZPPT Results
PTT-1, Manual Reactor Trip, Rev 14, 12/21/2010
PTT-2, Total Loss of Feedwater (Main and Emergency), Rev 16, 12/21/2010
Scenario-Based Testing for SES-161, 01/07/2011
Scenario Packages
SES-161, 01/11/2011
SES-134, 01/11/2011
SES-130, 01/19/2011
SES-054, 12/21/2009
SES-143, 01/17/2011
SES-135, 01/17/2011
SES-042, 12/21/2009
SES-030, 12/21/2009
JPM Packages
JPM-109, Transfer Auxiliary Steam from Main Steam to Unit 1 & 2 Steam, 01/19/2011
JPM-131, Place H2 Analyzer in Service, 01/19/2011
JPM-117, Perform the Low Vacuum Trip Test, 01/11/2011
JPM-120, Place EFP-2 in Standby Using AP-990 Enclosure 4, 01/17/2011
Attachment
4
Section 1R12: Maintenance Effectiveness
Procedures
ADM-NGGC-0203, Preventative Maintenance and Surveillance Testing Administration
ADM-NGGC-0107, Equipment Reliability Process Guidelines
EGR-NGGC-0156, Environmental Qualification of Electrical Equipment Important to Safety
Miscellaneous
Rosemount Report D8300040, Qualified Life Rosemount Transmitters
Calculation E89-0024, Determination of Qualified Life of Rosemount Transmitters
NCR data base review for failures of equipment associated with exceeding Vendor
Recommended Service Life
Nuclear Condition Reports
NCR 124994, EG System margin Recovery
NCR 432380, Functional failure of NI-2 during the period 8/11 - 8/12
NCR 426391, NI-2 is not indicating properly
NCR 434394, NI-15 indications on control board oscillating
NCR 434627, NI-2 is spiking
NCR 434911, NI-2 counts spiking during drain down
NCR 449395, NI-1 is drifting upwards
Maintenance Rule Reports
SE08-0010, January 30, 2008, Maintenance Rule Expert Panel Minutes
SE08-0015, March 26, 2008, Maintenance Rule Expert Panel Minutes
SE11-0011, January 24, 2011, Maintenance Rule Expert Panel Meeting Minutes
SE11-0007, Nuclear Instrumentation Source Range PMG Presentation to Remain in (a)(2)
Section 1R15: Operability Evaluations
NCR 436097, Evaluate MUV-164 for possible degradation
NCR 436075, Evaluate MUV-37 for possible degradation
Section 1R20: Refueling and Outage Activities
Procedures
AI-504, Guidelines for Cold Shutdown and Refueling
WCP-102, Outage Risk Management
Attachment
5
Section 4OA2: Problem Identification and Resolution
Nuclear Condition Reports
NCR 458032, Adverse trend in spent fuel pool FME station awareness
NCR 447796, Foreign material found in the spent fuel pool FMEA
NCR 447774, Pen used within the FMEA not documented in FME log
NCR 446420, Spent fuel pool FME log did not reflect material removed from area
NCR 434504, Spent fuel pool FMEA expanded without sufficient cleanliness inspection of
expanded area
NCR 435274, Plant personnel needlessly logging into spent fuel pool area log
Section 4OA5, Other Activities
Progress Energy Procedures
PT-178T, Special Procedure - Reactor Building Concrete Structural Integrity Test, Revision 0
SP-178T, Containment Leakage Test -Type "A" Including Liner Plate, Revision 0
Mistras Group, Inc Procedures.
Crystal River Unit 3 Tendon Retensioning Monitoring Procedure, Revision 4
Precision Surveillance Corporation (PSC) Field and Quality Control Procedures
3.0 Receiving, Handling and Storage, Revision 0
3.1, Equipment Proof Test, Revision 0
5.0, Tendon Initial Degreasing and Cap Removal, Revision 1
6.0, Tendon Detensioning/Removal for Possible Reuse, Revision 0
8.0, Plasma Cutting Tendon Detensioning, Revision 0
8.1, Ram Tendon Detensioning, Revision 2
9.0, Monitor Tendon Force (Lift-Offs), Revision 1
10.0, Tendon Removal, Revision 0
11.0, Tendon Void Cleaning, Revision 0
13.0, Tendon Installation, Revision 2
14.0, Tendon Field Anchor Head and Buttonheading Application, Revision 2
15.0, Tendon Restressing, Revision 3
15.1, Anchorage Inspection of Stressed Tendon, Revision 1
15.2, Bearing Plate Concrete Inspection, Revision 0
15.5, Additional Vertical Tendon Restressing, Revision 0
16.0, Grease Cap Replacement, Revision 0
17.0, Grease Replacement, Revision 1
Quality Assurance Procedures
10.0, Calibration of Measuring and Test Equipment, Revision 3
10.1, Verification of Calibrated Status of Hydraulic Pressure Gauges, Revision 0
Nuclear Condition Reports
378555 DBD 1/1, Containment, Has an Incorrect Value for Tendon Wire
422131 This NCR Tracks SGT NCR 154 Liner Plate Coatings Damage
422383 This NCR Tracks SGT IIRP 144 Dropped Hard Hat
Attachment
6
422487 This NCR Tracks SGT IIRP 145 Finger Injury
422488 This NCR Tracks SGT NCR 155 Tendon Buttonhead Discrepancies
440743 This NCR Is to track SGT NCR 148 Broken Tendon Wires
440778 Lost 0-2 Inch Dial Indicator
440785 This NCR Tracks SGT NCR 149/IIRP 135 Overlapping Shims
440833 Transposition Error In Attachment Z50 Caused Work Stoppage
441233 This NCR Tracks SGT NCR 150 Bent Test Wire
441239 This NCR Tracks SGT NCR 151 Liner Plate Coatings Defects
441332 This NCR Is to Track SGT IIRP 136 Dropped Object
441366 This NCR Tracks SGT NCR 152 Shim Stack Inconsistent
441394 SGT IIRP 138 Dropped Pendent Controller
441396 Bent Wire On Tendon 13H18 at Buttress 3
441453 This NCR Tracks SGT IIRP 140 Adverse Trend
441648 This NCR Tracks SGT IIRP 141 Dropped Object
442571 This NCR Tracks SGT NCR 156 Tendon Head Thread Issue
442706 This NCR Tracks SGT IIRP 146 Summary Of Work Related Issues
442711 This NCR Tracks SGT IIRP 147 Work Activities Ceased
442860 This NCR Tracks SGT DR 1024 Untimely Deviation Reporting
443077 This NCR Tracks SGT NCR 157 Missing Tendon Button Head
443290 This NCR Tracks SGT NCR 158 Tendon Wire Discrepancy
443343 This NCR Tracks SGT NCR 159 Broken Tendon Wire
443379 This NCR Tracks SGT NCR 160 Work Package Discrepancies
443424 This NCR Tracks SGT DR 26 Weld Rod Log Discrepancy
443487 Excess Leakage into the Tendon Access Gallery Sump
443529 Delay In Submitting SGT Qualification Cards
443563 Radio Repeater Power Loss Affected SGT Radio Communication
443566 SGT Workers Evacuated Tendon Galley Due To Exhaust Fumes
443587 FSP-2B Exhaust Disrupted Tendon Tensioning Work
443605 This NCR Tracks SGT NCR'S 162,163,164,165 Tendon 34V08 Issue
443606 This NCR Tracks SGT NCR 161 Missing Shim Spacer
443629 SGT Worker Put TLD Through NSOC X-Ray
443692 This NCR Tracks SGT DR 28 ILRT Configuration Control
Deficiencies
443749 Bottles with Improper Fluids Discovered in Tendon Galley
443961 This NCR Tracks SGT NCR 166 Tendon Lift Off Pressure Deltas
443962 This NCR Tracks SGT IIRP 154 Hydraulic Leak
443966 This NCR Documents PE Observation 47866 Lanyard Not on Tool
444707 This NCR Tracks SGT IIRP156/DR 29 Lift Off Pressure Delta
444726 This NCR Tracks SGT NCR 167 Bent Tendon Shims
444728 This NCR Tracks SGT NCR 168 Water in Tendon Can
444971 This NCR Tracks SGT NCR 169 Water In Tendon Can
444972 This NCR Tracks SGT NCR 170 Broken Tendon Wire
445558 This NCR Tracks SGT NCR 171 Broken Wires 13H32 51H32 51H34
445761 FSAR Description Not Implemented During Containment Repair
445762 Tendon Detensioning Basis Doc Does Not Exist For Surveillance
445853 This NCR Tracks SGT NCR 172-IIRP 161 Shim Spacing Delta
445982 This NCR Tracks SGT NCR 173/ IIRP 162 Hydraulic Leak
Attachment
7
446052 Laser Scan Data Is Outside of What's Expected
446084 This NCR Tracks SGT IIRP 163 Suspected Dropped Object
446225 This NCR Tracks SGT IIRP164 Dropped Object
446232 This NCR Tracks SGT IIRP 165 Broken Tendon Ram Gauge
446280 This NCR Tracks SGT IIRP 167 Vehicle Incident
446807 This NCR Tracks SGT NCR 174 Broken Wires 35H19 64H43 13H40
446808 This NCR Tracks SGT NCR 175 Stressing Sequence Discrepancy
447042 THIS NCR TRACKS SGT NCR 176 PASS 5 53H21 MISSING WIRE
447376 This NCR Tracks Tendon 34V08 Issue
447415 Cannot Couple on Tendon 34V08
447416 Raised Metal Noted on Tendon 34V01 Anchor Head
447417 Protruding Buttonhead on Tendon 34V13
448236 This NCR Tracks SGT NCR 180-34V08 Dome End Foreign Material
448468 This NCR Tracks SGT IIRP 176 Oil Spill on Dome
448504 Tendon Tensile Testing Delta between SGR and Design Basis
448535 This NCR Tracks SGT NCR 183 "As Found" Shim Gap Delta
448537 This NCR Tracks SGT NCR 181 Documentation Deficiency
448579 This NCR Tracks SGT NCR 182 Bearing Plate Concrete Gap Delta
448580 This NCR Tracks SGT NCR 184 Tendon 56V13 Damaged
Buttonhead
448581 This NCR Tracks SGT NCR 185 Bearing Plate Pitting 45V14
448621 Broken Wire On Tendon 45V01 SGT NCR 1186
448622 As Found Tendon Anchor Head Thread Damage 12V05,07,09,11,13
448639 SGT NCR 1188 Documents 23V20 Wire Delta
448640 NCR 1189 Pitting on 23V24 Bearing Plate
448659 This NCR Tracks SGT IIRP 178 Grease Hose Failure
449508 This NCR Tracks SGT IIRP 180 First Aid On Pinched Finger
449562 This NCR Tracks SGT NCR 190 Bearing Plate Concrete Gaps
450121 This NCR Tracks SGT NCR 191 Broken Tendon Wire 35H29
450133 This NCR Tracks SGT NCR 192 Bearing Plate Concrete Crack/Gap
450218 This NCR Tracks SGT IIRP 182 Harrington Hoist Not Working
450219 Tendon Gallery Bent Junction Box Cover SGT NCR 193
450591 This NCR Tracks SGT NCR 194 Tendon Elongation Delta Pass 8
451015 Worker Observed Standing on Platform 6 Hand Rail
451148 This NCR Tracks SGT NCR 197 As Found Flaking Coatings
451149 This NCR Tracks SGT NCR 198 Broken Wire 53H23 Shop End
451151 This NCR Tracks SGT NCR 201 Separated Existing Concrete
451166 This NCR Tracks SGT NCR 199 Broken Master Pressure Gauge
451167 This NCR Tracks SGT NCR 200 Gauge No "Post Use" Calibration
451306 This NCR Tracks SGT NCR 202 Protruding Wires Tendon 53H26
451412 Final Retensioning Using More Shims Than Anticipated
451425 This NCR Tracks SGT NCR 203 Pass 8 IWL Concrete Reject
451426 Strain Gage Alert Limit Exceeded By 0.3 Micro Strain
451471 This NCR Tracks SGT NCR 204 Bearing Plate Concrete Cracks
451549 This NCR Tracks SGT NCR 205 Broken Buttonhead 64H21
451632 This NCR Tracks SGT NCR 1206 Exposed Rebar Azimuth 200
Dome
Attachment
8
451727 This NCR Tracks SGT NCR 207 Bearing Plate Concrete Gaps 53H3
451730 This NCR Tracks SGT NCR 208 Bearing Plate Concrete Gaps
451732 This NCR TRACKS SGT IIRP 187 DROPPED OBJECT
451742 THIS NCR Tracks SGT NCR 209 Bearing Plate Concrete Gap 3H39
451900 This NCR Tracks SGT NCR 210 Bearing Plate Concrete Gaps
452030 IWL Inspection 51H41-35H41-53H41 Bearing Plate Concrete Crack
452032 This NCR Tracks SGT IIRP 190 Dropped Object
452178 This NCR Tracks SGT NCR 213 Anchorhead Delta 13H37 Field End
452188 IWL Inspection Bearing Plate Concrete Cracks 51H37 & 53H37
Drawings
75221-103, SIT Instrument Locations - Elevation 103-0
75521-135, SIT Instrument Locations - Elevation 135-0
75521-170, SIT Instrument Locations - Elevation 170-0
75521-205, SIT Instrument Locations - Elevation 205-0
75521-240, SIT Instrument Locations - Elevation 240-0
75521-250, SIT Instrument Locations - Elevation 250-0
75521-DOME, SIT Instrument Locations - From Reactor Building Dome
75521-EH, SIT Instrument Locations - Elevation 135-0
S-425-004, Revision 1, IWE/IWL Inspection Vertical Tendon Layout
S-425-005, Revision 1, IWE/IWL Inspection Hoop Tendon 13 Layout
S-425-006, Revision 1, IWE/IWL Inspection Hoop Tendon 42 Layout
S-425-007, Revision 2, IWE/IWL Inspection Hoop Tendon 53 Layout
S-425-008, Revision 1, IWE/IWL Inspection Hoop Tendon 64 Layout
S-425-009, Revision 1, IWE/IWL Inspection Hoop Tendon 51 Layout
S-425-010, Revision 1, IWE/IWL Inspection Hoop Tendon 62 Layout
Z08, Revision 9, Tendon Tensioning
Other Documents
Engineering Change 75221, Revision 3
Extensometer Installation Specifications and Guidance
Attachment