ML14178A535
| ML14178A535 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 06/27/2014 |
| From: | Nease R NRC/RGN-II/DRS/EB1 |
| To: | Batson S Duke Energy Corp |
| References | |
| IR-14-007 | |
| Download: ML14178A535 (29) | |
See also: IR 05000269/2014007
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
245 PEACHTREE CENTER AVENUE NE, SUITE 1200
ATLANTA, GEORGIA 30303-1257
June 27, 2014
Mr. Scott L. Batson
Site Vice President
Duke Energy Corporation
Oconee Nuclear Station
7800 Rochester Highway
Seneca, SC 29672-0752
SUBJECT:
OCONEE NUCLEAR STATION - NRC COMPONENT DESIGN BASES
INSPECTION REPORT 05000269/2014007, 05000270/20140007, AND
Dear Mr. Batson:
On May 9, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at
your Oconee Nuclear Station Units 1, 2 and 3. On June 18, 2014, the NRC team leader
discussed the results of this inspection with Oconee management and staff. Inspectors
documented the results of this inspection in the enclosed inspection report.
NRC inspectors documented two findings of very low safety significance (Green) in this report.
These findings involved violations of NRC requirements. Additionally, in this report, NRC
inspectors documented one Severity Level IV violation with no associated finding and a
licensee-identified violation, which was determined to be Severity Level IV. The NRC is treating
these violations as non-cited violations (NCV) consistent with Section 2.3.2.1 of the
If you contest the violations or significance of these NCVs, you should provide a response within
30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear
Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with
copies to the Regional Administrator, Region II; the Director, Office of Enforcement, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector
at the Oconee Nuclear Station.
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
Oconee Nuclear Station.
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 Public Document Room or from the Publicly Available Records (PARS) component of
NRC's Agencywide Document Access and Management System (ADAMS). ADAMS is
S. Batson
2
accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public
Electronic Reading Room).
Sincerely,
/RA/
Rebecca L. Nease, Branch Chief
Engineering Branch 1
Division of Reactor Safety
Docket Nos.: 05000269, 05000270, 05000287
License Nos.: DPR-38, DPR-47, DPR-55
Enclosure:
Inspection Report 05000269/2014007, 05000270/2014007, 05000287/2014007
w/Attachment: Supplementary Information
cc: Distribution via Listserv
__ML14178A535_______________
x SUNSI REVIEW COMPLETE x FORM 665 ATTACHED
OFFICE
DRS:RII
DRS:RII
DRS:RII
DRS:RII
DRS:RI
DRS:RII
DRS: RII
SIGNATURE
JAE1
REW1
WRM1
GLN
TLT3
NAME
J. Eargle
R. Williams
W. Monk
S. Sanchez
S. Pindale
G. Nicely
T. Tinkel
DATE
5/26/2014
5/21/2014
4/24/2014
4/23/2014
4/23/2014
4/23/2014
4/23/2014
E-MAIL COPY?
YES
NO
YES
NO YES
NO
YES
NO
YES
NO YES
NO
YES
NO
OFFICE
DRS:RII
DRS:RII
DRP:RII
RII:DRS
RII:DRS
RII:DRS
SIGNATURE
TNF1
RLN1
GJM1
JAE1
TNF1
RLN1
NAME
T. Fanelli
R. Nease
G. McCoy
J. Eargle
T. Fanelli
R. Nease
DATE
4/26/2014
6/11/2014
6/11/2014
6/23/2014
6/23/2014
6/27/2014
E-MAIL COPY?
YES
NO
YES
NO YES
NO
YES
NO
YES
NO YES
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Letter to Scott Batson from Rebecca L. Nease dated June 27, 2014
SUBJECT:
OCONEE NUCLEAR STATION - NRC COMPONENT DESIGN BASES
INSPECTION REPORT 05000269/2014007, 05000270/20140007, AND
Distribution:
RIDSNRRDIRS
PUBLIC
RidsNrrOconeeResource
Enclosure
U. S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos.:
50-269, 50-270, and 50-287
License Nos.:
Report Nos.:
05000269/2014007, 05000270/2014007, and 05000287/2014007
Licensee:
Duke Energy Carolinas, LLC
Facility:
Oconee Nuclear Station, Units 1, 2, and 3
Location:
7800 Rochester Highway
Seneca, SC 29672
Dates:
February 10, 2014 - March 28, 2014
April 28, 2014 - May 2, 2014
May 5, 2014 - May 9, 2014
Inspectors:
J. Eargle, Senior Reactor Inspector (Lead)
R. Williams, Senior Reactor Inspector
S. Pindale, Senior Reactor Inspector
S. Sanchez, Senior Emergency Preparedness Inspector
T. Fanelli, Reactor Inspector
W. Monk, Reactor Inspector
G. Nicely, Contractor (Electrical)
T. Tinkel, Contractor (Mechanical)
Approved by:
Rebecca L. Nease, Chief
Engineering Branch 1
Division of Reactor Safety
SUMMARY
IRs 05000269/2014-007, 05000270/2014-007 and 05000287/2014-007; 2/10/2014 - 5/9/2014;
Oconee Nuclear Station, Units 1, 2, and 3; Component Design Bases Inspection.
This inspection was conducted by a team of six Nuclear Regulatory Commission (NRC)
inspectors from Regions I and II, and two NRC contract personnel. Two Green non-cited
violations and one Severity Level (SL)-IV violation were identified. The significance of
inspection findings is indicated by their color (Green, White, Yellow, Red) using the NRC
Inspection Manual Chapter (IMC) 0609, Significance Determination Process, dated
June 2, 2011. Cross-cutting aspects are determined using IMC 0310, Aspects Within Cross
Cutting Areas, dated December 19, 2013. All violations of NRC requirements are dispositioned
in accordance with the NRCs Enforcement Policy, dated July 9, 2013. The NRC's program for
overseeing the safe operation of commercial nuclear power reactors is described in
NUREG-1649, Reactor Oversight Process, Revision 5.
NRC identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
Green: The team identified a Green non-cited violation of 10 CFR 50, Appendix B,
Criterion III, Design Control, for the licensees failure to ensure that at the worst-case
voltage, protective devices and thermal overload relays for safety-related loads would not
trip prior to and after the transfer to the emergency power source. This transfer occurs for a
sustained degraded voltage below the under voltage relay voltage settings for the duration
of the time delay setting or the manual actions credited. The licensee revised their voltage
calculations to account for previously unanalyzed loads. The licensee entered this issue
into its corrective action program as problem identification program (PIP) O-14-2280.
The team determined that the performance deficiency was more than minor because it was
associated with the Design Control 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 team identified that the voltages evaluated in the licensees analysis were non-
conservative and could result in lower unanalyzed voltages that could result in connected
safety-related loads stalling, becoming damaged, their protective devices tripping, or loads
such as battery chargers being below their minimum operating voltages. The team
determined that the finding was of very low safety significance (Green) because it was a
design deficiency that did not result in a loss of off-site power operability. The team
determined that no cross cutting aspect was applicable because this finding was not
indicative of current licensee performance. (Section 1R21.2.b.i)
Green: The team identified a Green non-cited violation (NCV), with two examples, of
10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to
correct conditions adverse to quality. Specifically, the licensee (1) failed to correct voltage
calculations for safety-related 4160 volt circuit breaker 125 volt-direct current control circuits
and (2) failed to correct voltage calculations for safety-related 120 volt alternating current
motor control center control circuits. The above issues were previously identified as NCV
05000269,270,287/2011010-04 and NCV 05000269,270,287/2011010-03, respectively.
3
The incomplete corrective actions were newly entered in the licensees corrective action
program as problem identification program (PIP) reports O-14-2781 and O-14-2811 to track
their completion.
The team determined that the performance deficiency was more than minor because it
affected the Equipment 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. The team
determined the finding was of very low safety significance (Green) because the inadequate
corrective actions did not result in losses of operability or function for either example. The
violation was assigned the cross-cutting aspect of Resolution in the area of Problem
Identification and Resolution because the licensee did not take effective corrective actions to
address issues in a timely manner. [P.3] (Section 1R21.2.b.ii)
SL-IV: The team identified a Severity Level IV non-cited violation of 10 CFR 50.71(e) for the
licensees failure to include in the latest Updated Final Safety Analysis Report (UFSAR)
changes made to the sites licensing bases with respect to station battery testing made
during the Technical Specification conversion to Integrated Technical Specifications.
Specifically, the UFSAR did not identify the standards by which the testing was conducted.
The licensee entered this issue into its corrective action program as problem identification
program report O-14-2338 and planned to include the omitted battery testing standards to
the UFSAR during an upcoming update cycle.
The team dispositioned the performance deficiency using the traditional enforcement
process because failing to update the UFSAR had the potential to adversely impact the
NRCs ability to perform its regulatory function. The performance deficiency was
characterized as a Severity Level IV violation in accordance with the NRC Enforcement
Policy, Section 6.1.d.3 as the lack of up-to-date information did not result in any
unacceptable change to the facility or procedures. In accordance with IMC 0612, Power
Reactor Inspection Reports, no cross-cutting aspects are assigned to traditional
enforcement violations. (Section 1R21.2.b.iii)
Licensee-Identified Violations
A violation of very low safety significance was identified by the licensee and has been reviewed
by the team. Corrective actions taken or planned by the licensee have been entered into the
licensees corrective action program. This violation and corrective action tracking numbers are
listed in Section 4OA7 of this report.
REPORT DETAILS
1.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
1R21 Component Design Bases Inspection (71111.21)
.1
Inspection Sample Selection Process
The team selected risk-significant components for review using information contained in
the licensees probabilistic risk assessment. In general, this included components that
had a risk achievement worth factor greater than 1.3 or Birnbaum value greater than 1E-
6. The sample included 18 components, two of which were associated with containment
large early release frequency (LERF), and five operating experience (OE) items.
The team performed a margin assessment and a detailed review of the selected risk-
significant components to verify that the design bases had been correctly implemented
and maintained. Where possible, this margin was determined by the review of the
design basis and the Updated Final Safety Analysis Report (UFSAR). This margin
assessment also considered original design issues, margin reductions due to
modifications, or margin reductions identified as a result of material condition issues.
Equipment reliability issues were also considered in the selection of components for a
detailed review. These reliability issues included items related to failed performance test
results, significant corrective action, repeated maintenance, maintenance rule status,
Regulatory Issue Summary 05-020 (formerly Generic Letter 91-18) conditions, NRC
resident inspector input regarding problem equipment, system health reports, industry
OE, and licensee problem equipment lists. Consideration was also given to the
uniqueness and complexity of the design, OE, and the available defense-in-depth
margins. An overall summary of the reviews performed and the specific inspection
findings identified is included in the following sections of the report.
.2
Component Reviews
a.
Inspection Scope
Components
Low Pressure (LP) Injection Motor-Operated Valves (MOV) LP-15 and LP-16
(Units 1, 2, and 3)
Condenser Circulating Water (CCW) MOVs -10, 11, 12, & 13 (Units 1, 2, and 3)
Turbine Driven Emergency Feedwater (TDEFW) Pumps (Units 1, 2, and 3)
TDEFW Steam Isolation Valves MS-93 (Units 1, 2, and 3)
TDEFW Steam Control Valves MS-87 (Units 1, 2, and 3)
Motor Driven Emergency Feedwater (MDEFW) Pumps (Units 1, 2, and 3)
Station Auxiliary Service Water (ASW) Pump
Keowee Governor Oil Pumps
Keowee Emergency Start Logic
Standby Shutdown Facility (SSF) Feed from Protected Service Water (PSW)
Switchgear
125 volt direct current (Vdc) Vital I&C Batteries (Units 1, 2, and 3)
CT6 and CT7 Transformer
5
SSF 600 volt alternating current (Vac) Motor Control Center 3XSF Breaker 3A
4160 Vac Breakers N1 and N2 (Unit 3)
SSF ASW Pump Motor
Diverse Scram System Channel 1 & 2 Pressure Transmitters 3RCPT0244
and 3RCPT0245
Components with LERF Implications
Turbine Bypass MOVs MS-19, 22, 28, & 31 (Units 1, 2, and 3)
Atmospheric Dump Valves MS-162, & 164 (Units 1, 2, and 3)
For the 18 components listed above, the team reviewed the plant technical specifications
(TS), UFSAR, design bases documents (DBDs), and drawings to establish an overall
understanding of the design bases of the components. Design calculations and
procedures were reviewed to verify that the design and licensing bases had been
appropriately translated into these documents. Test procedures and recent test results
were reviewed against DBDs to verify that acceptance criteria for tested parameters
were supported by calculations or other engineering documents, and that individual tests
and analyses served to validate component operation under accident conditions.
System modifications, vendor documentation, system health reports, preventive and
corrective maintenance history, and corrective action program documents were reviewed
(as applicable) in order to verify that the performance capability of the component was
not negatively impacted, and that potential degradation was monitored or prevented.
Maintenance Rule information was reviewed to verify that the component was properly
scoped, and that appropriate preventive maintenance was being performed to justify
current Maintenance Rule status. Component walkdowns and interviews were
conducted to verify that the installed configurations would support their design and
licensing bases functions under accident conditions and had been maintained to be
consistent with design assumptions. Documents reviewed are listed in the attachment to
this report.
Additionally, the team performed the following component-specific reviews:
The team observed a simulator scenario involving time critical actions for aligning
emergency core cooling system suction from the borated water storage tank to the
reactor building emergency sump (high pressure recirculation) to verify the required
operator actions could be accomplished within the required times and as relied upon
in design assumptions, and that the actions could be accomplished in accordance
with approved licensee procedures.
The team observed a simulator scenario involving operator actions to identify and
isolate a main steam line rupture following a reactor/turbine trip, and after emergency
operating procedure immediate actions had been performed, to verify the actions
could be accomplished as relied upon in design assumptions and in accordance with
approved licensee procedures.
The team performed table-top reviews, with a licensed operator, of several abnormal
and emergency procedures to better understand actions to be taken during a turbine
building flood and isolation of a faulted steam generator; then the team conducted in-
field walkdowns of these procedures to verify the actions could be accomplished
within the assumed timeframe, that there was sufficient guidance in the procedures
to properly complete the tasks, that equipment or tools necessary to assist in
accomplishing these tasks were available in the designated locations, and that the
6
areas requiring accessibility were indeed accessible; in addition, the team
interviewed operators qualified to these tasks to ensure their knowledge and training
was sufficient to successfully accomplish the tasks.
The team assessed the adequacy of the emergency power and DC control cabling
systems located in the concrete underground raceway to determine if any single
failure vulnerabilities existed.
b.
Findings
i
Failure to Evaluate the Under Voltage Relays at the Worst Case Minimum Drop Out Bus
Voltage
Introduction: The team identified a Green non-cited violation (NCV) of 10 CFR 50,
Appendix B, Criterion III, Design Control, for the licensees failure to ensure that at the
worst-case voltage, protective devices and thermal overload relays for safety-related
loads would not trip prior to and after the transfer to the emergency power source. This
transfer occurs for a sustained degraded voltage below the under voltage relay voltage
settings for the duration of the time delay setting or the manual actions credited.
Description: Licensee calculations OSC-2059, OSC-2060 and OSC-2061, documented
the bounding operating voltage requirements for Units 1, 2, and 3, respectively.
Appendix A to these calculations identified the minimum bus voltage levels to be 87.5%,
87.5% and 87.3% of the base 4160Vac bus voltage for Units 1, 2 and 3, respectively.
Each of these calculations analyzed the effect of operating safety-related loads below
the minimum acceptable voltages identified. The lowest voltage evaluated was at 84.7%
of the 4160Vac bus, which corresponded to 97% of the tap setting for the under voltage
relays. The licensee identified this as the must drop out point.
The team noted that the under voltage relays in use at Oconee were CV-7 inverse time
induction disk relays. The relay manufacturers documentation guaranteed that this type
of relay, on a decreasing voltage, would drop out at 97% of tap, but had no guaranteed
time for the drop out to complete. For decreasing voltages, it was not until the 90% of
tap setting (78.6% of bus voltage) or lower that the manufacturers documentation
guaranteed a specific time for the drop out to complete. Additionally, the team noted
that the calculations only evaluated operating equipment and did not account for system
transients or the effect of starting safety-related equipment when voltage was at the
must drop out point. Evaluation at the lower voltage could result in connected safety-
related loads stalling, becoming damaged, their protective devices tripping, or loads
such as battery chargers being below their minimum operating voltages for (1) the
degraded voltage time delay of 9+/-1 seconds for a degraded voltage and ES actuation
and (2) during manual actions for up to 12 minutes for a degraded voltage with no ES
actuation. In response, the licensee stated that although the manufacturer did not
provide a guaranteed drop out time at 97% of tap, specific testing would be performed
that verified that the under voltage relays would fully actuate within 16 seconds. The
licensee also performed an operability determination and determined that all energized
equipment could survive a system transient at the must drop out point and considered
the effects of starting safety-related equipment at the must drop out point.
7
Analysis: The licensees failure to ensure that at the worst-case voltage, protective
devices and thermal overload relays for safety-related loads would not trip prior to and
after the transfer to the emergency power for a sustained degraded voltage below the
under voltage relay voltage settings for the duration of the time delay setting or manual
actions credited was a performance deficiency and a violation of 10 CFR Part 50,
Appendix B, Criterion III. The team determined that the finding was more than minor
because it was associated with the Design Control 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 team identified that the voltages evaluated in the
licensees analysis were non-conservative and could result in lower unanalyzed voltages
that could result in connected safety-related loads stalling, becoming damaged, their
protective devices tripping, or loads such as battery chargers being below their minimum
operating voltages. The team used IMC 0609, Att. 4, Initial Characterization of
Findings, issued June 19, 2012, for Mitigating Systems, and IMC 0612, App. A, The
Significance Determination Process (SDP) for Findings At-Power, issued June 19,
2012, and determined the finding to be of very low safety significance (Green) because
the finding was a design deficiency that did not result in a loss of off-site power
operability. The team determined that no cross-cutting aspect was applicable because
this finding was not indicative of current licensee performance.
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires,
in part, that design control measures provide for verifying or checking the adequacy of
design and that design changes shall be subjected to design control measures
commensurate with those applied to the original design. Contrary to this, since
November 30, 2005, the licensee did not verify the adequacy of their design for safety-
related loads. Specifically, the licensee failed to verify that the connected safety-related
loads would not would not trip prior to and after the transfer to the emergency power
source for a sustained degraded voltage below the under voltage relay voltage settings
for the duration of the time delay setting or manual actions credited. This violation is
being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. The
violation was entered into the licensees corrective action program as PIP O-14-2280.
(NCV 05000269/2014007-01; 05000270/2014007-01; 05000287/2014007-01; Failure to
Evaluate the Under Voltage Relays at the Worst Case Minimum Drop Out Bus Voltage)
ii
Failure to Correct Issues with DC System Voltage Calculations and 120Vac Motor
Control Center (MCC) Control Circuit Calculations
Introduction: The team identified a Green NCV, with two examples, of 10 CFR 50,
Appendix B, Criterion XVI, for the licensees failure to correct a condition adverse to
quality. Specifically, the licensee (1) failed to perform all corrective actions identified to
correct voltage calculations for safety-related 4160V circuit breaker 125Vdc control
circuits and (2) failed to perform all corrective actions to correct voltage calculations for
safety-related 120Vac MCC control circuits. The above issues were previously identified
in NCV-05000269,270,287/2011010-04 and NCV-05000269,270,287/2011010-03,
respectively.
8
Description: The team identified the following deficiencies with the licensees corrective
actions:
Example 1 - Failure to Perform Corrective Actions to Correct Voltage Calculations
for Safety-Related 4160V Circuit Breaker 125Vdc Control Circuits
Licensee calculation OSC-4701, Operability Evaluation for PIR 0-092-0057,
determined the adequacy of 125Vdc control voltage to the ITE Type 5HK 4160V
breakers in the safety-related Keowee standby S and SK breaker switchgears (B1T
and B2T). Section 8 of this calculation determined that the required voltage to the
U1, U2 and U3 SK breakers was limited to between 58.8Vdc and 68.2Vdc, which
was inadequate to meet the 90Vdc minimum voltage rating for the close coils, as
specified by the breaker vendor. Section 10.14.6 of calculation OSC-4276, Oconee
125Vdc Vital Instrumentation and Control Voltage Adequacy, applied an alternate
acceptance criterion of 70Vdc for the close coils, based on testing documented in
Test Report (TR)-144, Oconee Emergency Power 5HK Switchgear Test. The
calculation determined that the lowest calculated required voltage at the close coils
of the S and SK circuit breakers was approximately 58.8Vdc, and concluded that the
breakers were capable of operation. The team noted that the testing documented in
OSC-4701 was a one-time field test of the actual breakers, and TR-144, performed
later, consisted of tests on only three specimens with the same nominal test
conditions and acceptance criteria specified in OSC-4701. The team also noted that
neither test controlled the environmental conditions such as aging or coil temperature
to determine whether the components would remain operable during design basis
conditions. Additionally, the team noted that the testing, performed under mild
conditions, indicated that some close coils failed from 45Vdc to 65Vdc. The team
determined that this variance in the predictability of operation under mild test
environments did not provide reasonable assurance of operation at voltages below
the 90Vdc minimum operating voltage and the more limiting design basis
environmental conditions.
The team noted that during the 2011 CDBI, a similar issue of concern was identified
for the SSF ITE 5HK 4160 breakers and was documented in PIP O-11-11438 and
dispositioned as NCV 05000269,270,287/2011010-04 Inadequate Control Circuit
Voltage Calculations. During that inspection the CDBI team identified other licensee
calculations that were susceptible to this issue including calculations OSC-4276 and
OSC-8113. OSC-4701 is used as an input to OSC-8113. In response to this
concern during the 2011 CDBI, the licensee initiated PIP O-11-11438 to update
design basis documentation to include testing criteria, to provide justification for
using the alternate acceptance criteria, and to update applicable procedures. During
the current inspection, the team noted that PIP O-11-11438 was closed and that
while it addressed concerns with testing the SSF breakers at lower voltages, it did
not address calculations OSC-4276 and OSC-8113, as identified in the associated
NCV, nor did it address the limiting design basis environmental conditions, such as
aging and coil temperature.
Example 2 - Failure to Perform Corrective Actions to Correct Voltage Calculations
for Safety-Related 120Vac Motor Control Center Control Circuits
Attachment 1 to licensee calculation OSC-5930, Unit 1 Motor Starter Circuit Voltage
and Fuse Adequacy Calculation, listed acceptance criteria for various types of
9
120Vac contactor coils used in 600V and 208V MCCs, ranging from 65% to 78.4% of
120V rated voltage. The criteria for contactor pickup voltage was based on various
tests and was lower than the criteria specified in applicable National Electrical
Manufacturing Association (NEMA) standards (NEMA ICS-2) of 85%. During the
2011 CDBI, in NCV 05000269,270,287/2011010-03, Failure to Perform Adequate
Calculations to Support Keowee Voltage Trip Setpoints, the team identified the
following concerns:
For Sylvania TM starters only two specimens each of size 1 and 2 were tested,
providing an inadequate basis for the rating.
For Joslyn Clark and Cutler Hammer contactors the calculation took credit for
Control Power Transformer boost (approximately 2-4%) that had already been
credited in tests.
Tests were conducted on contactors at shop ambient temperature (cold coil).
Contactors may have been required to operate in service with hot coils. This
could have raised the pickup voltage by approximately 4%.
The acceptance criteria in the calculation did not provide margin over test criteria
to account for degradation over the service life of the contactors. Contactors
were not periodically tested to confirm low pickup voltage capability.
The calculation contained incomplete or obsolete information (e.g. contactors
that have been replaced and test reports missing).
In response, the licensee initiated PIP O-11-11440 to revise the instrument
procedures and validate the previous test values. Additionally, the PIP contained
actions to develop a periodic testing plan that would validate the minimum pickup
voltages being used for each type of starter/contactor and ensure they have not
degraded, and to generate action requests to perform the new periodic testing on the
starters/contactors. During the current inspection, it was identified that PIP O-11-
11440 was closed on 10/9/2013 with some of the identified corrective actions not
implemented and some that did not fully address each issue identified above.
Specifically, the scope of the periodic testing was identified to be random and not
comprehensive, the testing did not account for design basis environmental conditions
(e.g. testing the contactors with hot coils), and the testing values specified in the
implementing procedures had not been verified to correspond to those used by the
latest calculations or documented in a design deliverable document.
Analysis: The licensees failure to perform corrective actions to correct voltage
calculations for safety-related 4160V circuit breaker 125Vdc control circuits and to
correct voltage calculations for safety-related 120Vac MCC control circuits as required
by 10 CFR Part 50, Appendix B, Criterion XVI, was a performance deficiency. The team
determined that the performance deficiency was more than minor because it affected the
Equipment 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,
failing to account for aging and environmental effects could negatively impact the
reliability of the affected safety-related electrical components. The team used IMC 0609,
Att. 4, Initial Characterization of Findings, issued June 19, 2012, for Mitigating
Systems, and IMC 0612, App. A, The Significance Determination Process (SDP) for
Findings At-Power, issued June 19, 2012, and determined the finding to be of very low
safety significance (Green) because the incomplete corrective actions did not result in
10
losses of operability or function for any of the examples. The violation was assigned the
cross-cutting aspect of Resolution, in the area of Problem Identification and Resolution,
because the licensee did not take effective corrective actions to address the issues in a
timely manner. [P.3]
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions,
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, since October 9, 2013, the licensee failed to correct conditions adverse to
quality. Specifically, the licensee (1) failed to correct voltage calculations for safety-
related 4160V circuit breaker 125Vdc control circuits and (2) failed to correct voltage
calculations for safety-related 120VAC MCC control circuits. This violation is being
treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. The
violation was entered into the licensees corrective action program as PIPs O-14-2781
and O-14-2811. (NCV 05000269/2014007-02; 05000270/2014007-02; 05000287/2014007-02; Failure to Correct Issues with the DC System Testing and
120Vac Motor Control Center Control Circuits)
iii
Failure to Update the UFSAR with Current Battery Testing Standards
Introduction: The team identified a Severity Level IV NCV of 10 CFR 50.71(e) for the
licensees failure to include in the latest Updated Final Safety Analysis Report (UFSAR)
changes made to the sites licensing basis with respect to station battery testing made
during the Technical Specification conversion to Integrated Technical Specifications.
Specifically, the UFSAR did not identify the standards by which the testing is conducted.
Description: During a review of battery test procedures the team determined that
Oconee had not established design basis requirements in the UFSAR to support the
licensing basis requirements in Technical Specification 5.5.20, Battery Discharge
Testing Program. The licensee documented this issue in their corrective action
program with PIP O-14-02338, which stated that during the TS conversion to Integrated
Technical Specifications, an apparent change was made to the plants licensing basis
with respect to station battery testing. This licensing basis change was not subsequently
reflected in the UFSAR as required by 10 CFR 50.71(e). Specifically, the UFSAR did
not identify the standard(s) by which testing is conducted.
Analysis: The licensees failure to update the UFSAR with current battery testing
standards, as required by 10 CFR 50.71(e), was a performance deficiency. The team
dispositioned the performance deficiency using the traditional enforcement process
because failing to update the UFSAR had the potential to adversely impact the NRCs
ability to perform its regulatory process. The performance deficiency was more than
minor because the failure to provide complete licensing and design basis information in
the UFSAR could result in either the licensee making an inappropriate licensing
interpretation or the NRC making an inappropriate regulatory decision based on
incomplete information in the UFSAR. The performance deficiency was characterized
as a Severity Level IV violation in accordance with the NRC Enforcement Policy,
Section 6.1.d.3 as the lack of up-to-date information did not result in any unacceptable
change to the facility or procedures. In accordance with IMC 0612, Power Reactor
Inspection Reports, issued January 24, 2013, there are no cross-cutting aspects
assigned to traditional enforcement violations.
11
Enforcement: Title 10 CFR 50.71(e) requires in part, that licensees shall periodically
update the Final Safety Analysis Report (FSAR), originally submitted as part of the
application for the operating license, to assure that the information included in the report
contains the latest information developed. The submittal shall include the effects of all
changes made in the facility or procedures as described in the FSAR. Contrary to the
above, since December 16, 1998, the licensee failed to update the UFSAR to assure
that the information included in the report contained the latest information developed.
Specifically, the licensee failed to identify the current standards by which the station
battery testing was conducted. The failure to update the UFSAR as required
by 10 CFR 50.71(e) was characterized as a Severity Level IV violation. This violation is
being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. The
violation was entered into the licensees corrective action program as PIP O-14-2338.
(NCV 05000269/2014007-03; 05000270/2014007-03; 05000287/2014007-03; Failure to
Update the UFSAR with Current Battery Testing Standards)
iv
(Opened) Degraded Voltage Relay Scheme
Introduction: The team identified an unresolved item (URI) to determine whether a
performance deficiency exists with respect to the licensees degraded voltage relay
scheme.
Description: The team identified that the licensees degraded voltage relays did not
monitor the safety-related 4.16kV buses, but rather they monitored the switchyard 230kV
Yellow bus. This resulted in a lack of degraded voltage protection whenever the 4.16kV
safety-related buses were not being fed through the start-up transformers. During
normal power operation, the 4.16kV safety-related buses were supplied from the unit
auxiliary transformers. Additionally, for degraded voltage detected on the 230kV
switchyard Yellow bus with no accident signal present, the degraded voltage relay alarm
in the main control room would have only resulted in manual actions to resolve the
degraded voltage condition or to disconnect from the degraded source. It was estimated
that the manual actions could take as long as 12 minutes to resolve the degraded
voltage condition. The use of degraded voltage relays only on the 230kV switchyard
Yellow bus and the use of manual actions for a degraded voltage condition appeared to
be contrary to the design criteria for degraded voltage protection stated in an NRC letter
to the licensee dated June 3, 1977 and NRC Regulatory Issue Summary 2011-12.
Lastly, the team identified that Oconee currently credits operation of the loss-of-voltage
relays monitoring the 4.16kV main feeder buses to disconnect from offsite power on a
loss of voltage condition and subsequent re-connection to Keowee Hydro to meet the
UFSAR Chapter 15 plant accident analyses. However, the loss of voltage relay
setpoints and associated time delays were not included in the plant TS. This appeared
to be contrary to 10 CFR 50.36(c)(2)(ii)(C) Criterion 3.
The team determined that consultation with the Office of Nuclear Reactor Regulation
was warranted for the NRC to determine: (1) whether Oconees existing licensing and
design bases are adequate and meet all NRC regulations and requirements with their
current degraded voltage relays design and off-site/station electric power system design,
(2) whether the automatic actions for the loss-of-voltage relays meet the intent of the
degraded voltage relays, and (3) whether the current plant TS meet the requirements of
10 CFR 50.36(c)(2)(ii)(C) which state, in part, that a TS limiting condition for operation of
a nuclear reactor must be established for a structure, system, or component that is part
of the primary success path and which functions or actuates to mitigate a design basis
accident or transient that either assumes the failure of or presents a challenge to the
12
integrity of a fission product barrier. The licensee entered this issue into their corrective
action program as PIP O-14-2034. This issue is being tracked as URI 05000269/2014007-04, 05000270/2014007-04, 05000287/2014007-04, Degraded
Voltage Relay Scheme.
v. (Opened) Potential Unanalyzed Condition Associated with Emergency Power System
Introduction: The team identified a URI to determine whether a performance deficiency
exists related to the configuration of electrical cabling in the underground concrete
raceway. Specifically, the team was concerned that short circuits and/or ground faults in
the cabling could potentially impact the functionality of the emergency power system
which is required to mitigate certain design basis events.
Description: During a review of Oconees engineered safeguards protection system
(ESPS) emergency power start control for the KHUs, the team noted that the 125Vdc
control cables for train A of the ESPS and cables for supervisory control of both KHUs
were recently modified. The team also noted that these 125Vdc control cables were
installed in the same underground concrete raceway systems as the 4160Vac auxiliary
power cables, 13.8kVac power cables for both emergency power and protected service
water (PSW), and were in close proximity to these power cables. The team was
concerned that a short circuit (which the licensee considered outside their design basis)
in the 13.8kVac cables could induce voltage and currents in the dc control system which
could potentially impact the functionality of the emergency power system which is
required to mitigate certain design basis events. A similar issue exists in Manhole 6 of
the PSW underground raceway where the new power supply to the PSW (adjacent to
the 125Vdc control emergency power system) could short circuit or fault to ground. The
licensee had not performed an analysis to determine the effects of such failures on the
ability of the emergency power system to perform its safety function, thus the team
questioned whether the plant was in an unanalyzed condition. Although the licensee did
not agree that these failures were part of their licensing basis, they reported this as an
unanalyzed condition to the NRC in accordance with 10 CFR 50.73(a)(2)(ii)(B) in
Licensee Event Report 269/2014-01. In response to the teams concerns, the licensee
entered this issue into their corrective action program, and performed immediate and
prompt determinations of operability in which they concluded a reasonable expectation
of operability exists.
The team has requested assistance from subject matter experts in the Office of Nuclear
Reactor Regulation via a Task Interface Agreement1 to review the emergency power
system licensing basis to determine the acceptability of the licensees design. If the
design is found to be noncompliant with the licensing basis, the licensee will be required
to implement corrective actions to restore compliance.
This issue is being tracked as URI 05000269/2014007-05, 05000270/2014007-05, 05000287/2014007-05, Potential Unanalyzed Condition Associated with Emergency
Power System.
1 A Task Interface Agreement is a request for technical assistance to the Office of Nuclear Reactor Regulation
(NRR) on subjects within the scope of NRRs mission. In this case, there is a lack of clarity on whether the
licensees current design complies with the licensees licensing basis and NRR is being asked to establish the agency
position.
13
.3
Operating Experience
a.
Inspection Scope
The team reviewed five operating experience issues for applicability at Oconee Nuclear
Station, Units 1, 2, and 3. The team performed an independent review of these issues
and, where applicable, assessed the licensees evaluation and dispositioning of each
item. The issues that received a detailed review by the team included:
NRC Information Notice (IN) 2013-14, Potential Design Deficiency in Motor-
Operated Valve Control Circuitry, dated August 23, 2013
NRC IN 2013-05, Battery Expected Life and Its Potential Impact on Surveillance
Requirements, dated March 19, 2013
NRC IN 2012-16, Preconditioning of Pressure Switches Before Surveillance
Testing, dated August 29, 2012
NRC IN 2012-14, Motor-Operated Valve Inoperable Due To Stem-Disc
Separation, dated July 24, 2012
NRC IN 2012- 06, Ineffective Use of Vendor Technical Recommendations,
dated April 24, 2012
b.
Findings
No findings were identified.
4OA7 Licensee-Identified Violations
The following Severity Level IV violation was identified by the licensee and is a violation
of NRC requirements which met the criteria of the NRC Enforcement Policy for being
dispositioned as a NCV.
10 CFR 50.71(e) requires, in part, that each person licensed to operate a nuclear
power reactor, shall update periodically, the FSAR originally submitted as part of the
application for the license, to assure that the information included in the report
contains the latest information developed. This submittal shall include the effects of
all changes made in the facility or procedures as described in the FSAR. Contrary
to the above, since December 6, 2012, after updating the UFSAR to reflect the new
licensing basis under NFPA-805, several items applicable to the Fire Protection
System were incorrectly removed. Traditional enforcement was applicable because
the violation could impact the regulatory process, and was evaluated using
the NRCs Enforcement Policy. This violation was determined to be a Severity
Level IV violation because the lack of up-to-date information did not result in an
unacceptable change to the facility or procedures. This violation was documented in
the licensees corrective action program as PIP O-13-09302.
14
4OA6 Meetings, Including Exit
On March 20, 2014, the team leader presented the inspection results to Mr. Scott Batson
and other members of the licensees staff. On March 28, 2014, May 9, 2014, and June
18, 2014, the team leader discussed the results of the inspection with Oconee
management and other members of the licensees staff. The team verified that no
proprietary information was documented in this report.
ATTACHMENT: SUPPLEMENTARY INFORMATION
Attachment
SUPPLEMENTARY INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
K. Alter, Regulatory Affairs Manager
K. Anderson, BOP Supervisor
S. Batson, Site Vice President
V. Bowman, Design Engineering Manager
J. Brady, Regulatory Affairs
E. Burchfield, Engineering Manager
T. Patterson, Safety Assurance Manager
R. Price, Design Engineering Manager
J. Smith, Regulatory Affairs
C. Wasik, Regulatory Compliance Manager
N. Watson, IST Program Coordinator
NRC personnel
E. Crowe, Oconee Senior Resident Inspector
G. Matharu, Sr. Electrical Engineer, Office of Nuclear Reactor Regulation (NRR)
R. Mathew, Team Leader, NRR
J. Zimmerman, Branch Chief, NRR
LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED
Opened and Closed
05000269, 270, 287/2014007-01
Failure to Evaluate the Under Voltage Relays
at the Worst Case Minimum Drop Out Bus
Voltage [Section 1R21.2.b.i]
05000269, 270, 287/2014007-02
Failure to Correct Issues with DC System
Voltage Calculations and 120Vac Motor
Control Center (MCC) Control Circuit
Calculations [Section 1R21.2.b.ii]
05000269, 270, 287/2014007-03
SL-IV
Failure to Update the UFSAR with Current
Battery Testing Standards [Section 1R21.2.b.iii]
Opened
05000269, 270, 287/2014007-04
Degraded Voltage Relay Scheme [Section
1R21.2.b.iv]
05000269, 270, 287/2014007-05
Potential Unanalyzed Condition Associated
with Emergency Power System [Section
1R21.2.b.v]
LIST OF DOCUMENTS REVIEWED
Procedures
AD-EG-ALL-1104, Obsolescence Program, Rev. 0
AD-EG-ALL-1105, Critical Spare Program, Rev. 0
AP/0/A/1700/006, Natural Disaster, Rev. 2
AP/0/A/1700/025, Standby Shutdown Facility Emergency Operating Procedure, Rev. 60
AP/1/A/1700/010, Turbine Building Flood, Rev. 8
AP/1/A/1700/028, ICS Instrument Failures, Rev. 20
AP/1/A/1700/034, Degraded Grid, Rev. 11
AP/1/A/1700/039, Unintentional Boration, Rev. 002
AP/2/A/1700/026, Loss of Decay Heat Removal, Rev. 23
AP/3/A/1700/010, Turbine Building Flood, Rev. 8
AP/3/A/1700/013, Dam Failure, Rev. 20
EM 5.1, Engineering Emergency Response Plan, Rev. 32
EP/1/A/1800/001, Enclosure 5.1, Engineered Safeguards Actuation, Rev. 39
EP/1/A/1800/001, Enclosure 5.10, Station ASW Pump Alignment, Rev. 39
EP/1/A/1800/001, Enclosure 5.11, Reactor Coolant System Boration, Rev. 39
EP/1/A/1800/001, Enclosure 5.12, ECCS Suction Swap to RBES, Rev. 39
EP/1/A/1800/001, Enclosure 5.24, Operation of the Atmospheric Dump Valves, Rev. 39
EP/1/A/1800/001, Enclosure 5.31, Temporarily Charging the HPSW System, Rev. 39
EP/1/A/1800/001, Enclosure 5.34, Aligning SSF-ASW for Steam Generator Feed, Rev. 39
EP/1/A/1800/001, Enclosure 5.36, Equipment Alignment for Plant Shutdown, Rev. 39
EP/1/A/1800/001, Enclosure 5.4, Makeup to the Borated Water Storage Tank, Rev. 39
EP/1/A/1800/001, Enclosure 5.42, Alignment of EFM Pump to Feed Steam Generators, Rev. 39
EP/1/A/1800/001, Enclosure 5.5, Pressurizer and Letdown Storage Tank Level Control, Rev. 39
EP/1/A/1800/001, Enclosure 5.8, Feeding Steam Generators with Station ASW, Rev. 39
EP/1/A/1800/001, Enclosure 5.9, Extended Emergency Feedwater Operation, Rev. 39
EP/1/A/1800/001, HPI Cooldown, Rev. 39
EP/1/A/1800/001, LOCA Cooldown, Rev. 39
EP/1/A/1800/001, Rule 2, Loss of Subcooling Margin, Rev. 39
EP/1/A/1800/001, Rule 3, Loss of Main or Emergency Feedwater, Rev. 39
EP/1/A/1800/001, Rule 5, Main Steam Line Break, Rev. 39
EP/1/A/1800/001, Rule 7, Steam Generator Feed Control, Rev. 39
EP/1/A/1800/001, Turbine Building Flood, Rev. 39
EP-1-A-1800-001, Encl 5.8- 5.10, Feeding SGs With Station ASW, 0
IP/0/A/2007/005, PSW NLI Transformer Inspection and Maintenance, Rev. 1
IP/1/A/0400/034, KHU-1 Governor Oil System Pressure and Level Instrument Calibration,
Rev. 19
IP/3/A/0200/042, RCS ICCM-86 System RVLIS Instrument Calibration, Rev. 48
MP/0/A/2000/072, Keowee Hydro Station Pump - Governor Oil - Disassembly, Repair and
Assembly, Rev. 8
MP/0/A/2000/075, KHS Oil Sampling, Rev. 6
MP/1/A/2200/001, KHU-1 Governor Oil Pump Assemblies Inspection and Maintenance, Rev. 12
NAP000LW, Duke Energy Nuclear Scaffold Manual, Rev. 4
NSD 106, Configuration Management, Rev. 7
NSD 204, Operating Experience Program, Section 204.6.6, Rev. 16
NSD 229, Evaluation & Reporting of Deviations and Noncompliance per 10CFR21, Rev. 5
NSD 319, Vendor Technical Information Program, Rev. 4
NSD 408, Testing, Rev. 7
OP/0/A/1102/024, Plant Assessment/Alignment Following Major Site Damage, Rev. 36
3
OP/0/A/1102/025, Cooldown Following Major Site Damage, Rev. 25
OP/0/A/2000/027, KHU-1 Governor Oil Pumps, Rev. 8
OP/0/A/2000/043, KHS Shift Turnover and Rounds, Rev. 39
OP/0/A/6100/016, Alarm Response Guide SA-16, Rev. 11
OP/1/A/1102/004, Operation at Power, Rev. 139
OP/1/A/1102/008, On-Line Valve Lineup for MOV Maintenance, Rev. 39
OP/1/A/1102/010, Controlling Procedure for Unit Shutdown, Rev. 215
OP/1/A/1104/004, Low Pressure Injection System, Rev. 145
OP/1/A/1106/030, Identification of Failed Steam Generator Tubes, Rev. 20
OP/1/A/2000/102, KHU-1 Alarm Response Guide 1SA-2, Rev. 8
OP/3/A/1104/012E, Isolation and Reflooding of CCW Inlet Piping, Rev. 17
OTP 4116.2, Nuclear Equipment Operator Requalification, Rev. 19
OTP S601.0, Attachment 9-3, Job Performance Measure Evaluation Record, Rev. 16
PT/0/A/0251/010, Auxiliary Service Water Pump Test, Rev. 60
PT/0/B/0120/032, Field Equipment and Procedures Surveillance, Rev. 43, 10/4/2013
PT/0/B/0120/032, Field Equipment and Procedures Surveillance, Rev. 43, 11/22/2013
PT/0/B/0120/032, Field Equipment and Procedures Surveillance, Rev. 43, 12/20/2013
PT/0/B/0120/032, Field Equipment and Procedures Surveillance, Rev. 43, 12/22/2013
PT/2/A/0600/012, Turbine Driven Emergency Feedwater Pump Test, Rev. 91
PT/2/A/0600/28, 2MS-93 Nitrogen Supply Leakage Test, Rev. 0
PT/3/A/0600/013, Motor Driven Emergency Feedwater Pump Test, Rev. 61
RE-3.03, MCC Breaker and Overload Heater Selection, Rev. 4
SCD 282, Nuclear Supply Chain Process Manual - Shelf Life, Rev. 17
SCD 410, Nuclear Supply Chain Process Manual - Receiving, Rev. 17
Unit 1 EOP Enclosures 5-26, Manual Start of TDEFW Pump, Rev. 39
Unit 1 EOP Enclosures 5-27, Alternate Methods for Controlling EFDW Flow, Rev. 39
Drawings
29411642-NP1-1, Transformer CT6 Nameplate, Rev. 3
29411642-NP2-1, Transformer CT7 Nameplate, Rev. 3
29411642-SWD1-1, Transformer CT6 Schematic and Wiring Diagram, Rev. 0
29411642-SWD2-1, Transformer CT7 Schematic and Wiring Diagram, Rev. 0
41239, Atmospheric Steam Vent Valve, Rev. G
67247-1, Drag Valve 8x10, Globe, 600 ANSI 12 Inch Outlet Turbine Bypass, Rev. F
672471-1, 12 Inch Outlet Turbine Bypass Valve, 11/8/2006
B-16419, Bingham Pump 7 Stage Double Suction Double Volute Pump Type MSD-D, Rev. 0
B-27308, Bingham-Willamette 10 Stage - Single Suction Double Volute Pump Type MSD,
Rev. 0
C-71339, 96 Rubber Seated BFV with Limitorque Actuator and Extension Bonnet, Rev. 0
C-71500, Materials List for 96 R1A w/Extension Bonnet Access Opening, Rev. 1
CC02525, General Assembly (Pacific) Bolted Bonnet Flex-Wedge Weld End Gate Valve
w/SMB-00, Sht. 1, Rev. I
K-422A-23, Governor Oil Pump A Pressure, Rev. 3
K-422A-24, Governor Oil Pump C Pressure, Rev. 2
K-422A-25, Governor Oil Pump C Pressure, Rev. 2
KEE-107, Elementary Diagram Governor Oil Pump No. 1A, 1B, and 1C, Rev. 9
KEE-111-A, Elementary Diagram Turbine and Governor Systems Startup Control, Rev. 3
KFD-105A-1.1, Flow Diagram of Governor Oil System, Rev. 7
KM-200-0084, Pump Arrangement, Rev. 1
KM-200-0085, Hydraulic Schematic KHU2, Rev. D1
O-0702-B, SSF 4160 and 600 V Essential Load Centers, Rev. 26
4
O-0729-A, ATWS Mitigation Systems (AMSAC/DSS) Control PNL 1ATWSCP, Rev.1
O-1422M-33, Instrument Details, Turbine Bypass Control Valve 2MS19, Rev. 6
O-1422M-33-01, Instrument Details, Turbine Bypass Control Valve 2MS22, Rev. 1
O-1422M-33-02, Instrument Details, Turbine Bypass Control Valve 2MS28, Rev. 1
O-1729-A, ATWS Mitigation Systems (AMSAC/DSS) Control PNL 2ATWSCP, Rev. 2
O-2729-A, ATWS Mitigation Systems (AMSAC/DSS) Control PNL 3ATWSCP, Rev. 2
O-6700, Main PSW Switchgear One Line Diagram, Rev. 3
OEE-117-93-OA, SSF 4160V SWGR OTS1 ASW Pump Motor Feeder, Rev. 0
OEE-136-6, CCW System CCW Pump C Discharge Valve, Rev. 9
OEE-352-3, Elementary Diagram LPI Pump 3A Discharge Valve, Rev. 3
OEE-352-4, Elementary Diagram LPI Pump 3B Discharge Valve, Rev. 3
OFD-102A-1.2, Flow Diagram Unit 1 LPI System, Rev. 54
OFD-102A-2.2, Flow Diagram Unit 2 LPI System, Rev. 47
OFD-102A-3.2, Flow Diagram Unit 3 LPI System, Rev. 44
OFD-121A-1.8, Flow Diagram of Condensate System, Rev. 25
OFD-121B-3.3, Flow Diagram of Feedwater System, Rev. 29
OFD-121D-1.1, Flow Diagram of Emergency Feedwater System, Rev. 37
OFD-121D-1.2, Flow Diagram of Emergency Feedwater (Auxiliary Service Water), Rev. 24
OFD-121D-2.1, Flow Diagram of Emergency Feedwater System, Rev. 38
OFD-121D-3.1, Flow Diagram of Emergency Feedwater System, Rev. 44
OFD-122A-1.1, Flow Diagram of Unit 1 Main Steam System (Main Steam Headers 1A and 1B),
Rev. 24
OFD-122A-1.2, Flow Diagram of Unit 1 Main Steam System (Turbine Bypasses), Rev. 19
OFD-122A-1.4, Flow Diagram of Main Steam System Emergency FDW Pump Turbine Steam
Supply & Exhaust, Rev. 24
OFD-122A-2.1, Flow Diagram of Main Steam System, Rev. 24
OFD-122A-2.1, Flow Diagram of Unit 2 Main Steam System (Main Steam Headers 2A and 2B),
Rev. 24
OFD-122A-2.2, Flow Diagram of Unit 2 Main Steam System (Turbine Bypasses), Rev. 18
OFD-122A-2.4, Flow Diagram of Main Steam System Emergency FDW Pump Turbine Steam
Supply & Exhaust, Rev. 23
OFD-122A-3.1, Flow Diagram of Main Steam System, Rev. 32
OFD-122A-3.1, Flow Diagram of Unit 3 Main Steam System (Main Steam Headers 3A and 3B),
Rev. 32
OFD-122A-3.2, Flow Diagram of Unit 3 Main Steam System (Turbine Bypasses), Rev. 18
OFD-122A-3.4, Flow Diagram of Main Steam System Emergency FDW Pump Turbine Steam
Supply & Exhaust, Rev. 27
OFD-133A-1.1, Flow Diagram of Unit 1 CCW System (CCW Intake Pumps Discharge), Rev. 31
OFD-133A-2.1, Flow Diagram of Condenser Circulating Water System (CCW Intake Pumps
Discharge), Rev. 32
OFD-133A-2.1, Flow Diagram of Unit 2 CCW System (CCW Intake Pumps Discharge), Rev. 32
OFD-133A-3.1, Flow Diagram of Unit 3 CCW System (CCW Intake Pumps Discharge), Rev. 40
OM 206.A-0001 001, Bingham Pump 4x8x10-1/2 MSD 0, Rev. 2
OM 206.A-0006.1 001, Unit 1 Turbine-Driven EFW Pump Curve 26983-2 (S/N 280065),
10/17/2001
OM 206.A-0007.1 001, Unit 2 Turbine-Driven EFW Pump Curve 26984-2 (S/N 280066),
1/17/2002
OM 206.A-0008.1 001, Unit 3 Turbine-Driven EFW Pump Curve 26985-2 (S/N 280067),
10/17/2001
OM 206-0037 001, 3B Motor-Driven EFW Pump Curve 35816 (S/N 14210669), 5/10/1990
OM 206-0038 001, 2B Motor-Driven EFW Pump Curve 35815 (S/N 14210668), 12/18/1990
5
OM 206-0039 001, 2A Motor-Driven EFW Pump Curve 35814 (S/N 14210667), 11/30/1990
OM 206-0040 001, 3A Motor-Driven EFW Pump Curve 35813 (S/N 14210666), 5/10/1990
OM 206-0041 001, 1B Motor-Driven EFW Pump Curve 35812 (S/N 220057)
OM 206-0042 001, 1A Motor-Driven EFW Pump Curve 35811 (S/N 220056)
OM 206-0045 001, 2A Motor-Driven EFW Pump Curve (modified) 35814-A (S/N 14210667),
12/18/1990
OM 206-0046 001, 2B Motor-Driven EFW Pump Curve (modified) 35814-A (S/N 14210668),
9/16/1991
OM 206-0047 001, 3A Motor-Driven EFW Pump Curve (modified) 35813-A (S/N 14210666),
3/15/1991
OM 206-0048 001, 3B Motor-Driven EFW Pump Curve (modified) 35816-A (S/N 14210669),
3/15/1991
OM 208.-0344.002, ASW Pump Curve C0869073B (S/N 0869-73), Rev. B
OM 251.-0762 001, Drag Valve 6x6 Globe, Rev. B
OM 251-0793.001, Fisher 6 Body 50 Actuator 657 ED Diaphragm Actuated Control Valve,
Rev. B
ONTC-0-127C-0001-001, ONS Units 1, 2, and 3 MS-93 Nitrogen Manifold Leakage Test
Acceptance Criteria, Rev. 1
ONTC-1-121D-0001-001, ONS Unit 1 Test Acceptance Criteria for MDEWF Pumps 1A and 1B,
Rev. 6
ONTC-1-121D-0002-001, ONS Unit 1 Test Acceptance Criteria TDEFW Pumps, Rev. 4
ONTC-1-121D-0003-001, ONS Unit 1 Test Acceptance Criteria for TDEFW Pump Minimum
Flow Line Orifice, Rev. 1
ONTC-2-121D-0001-001, ONS Unit 2 Test Acceptance Criteria for MDEWF Pumps 2A and 2B,
Rev. 6
ONTC-2-121D-0003-001, ONS Unit 2 Generate New Performance Curves for MDEFW Pumps
2A and 2B, Rev. 0
ONTC-3-121D-0001-001, ONS Unit 3 Test Acceptance Criteria for MDEWF Pumps 3A and 3B,
Rev. 7
ONTC-3-121D-0003-001, ONS Unit 3 Test Acceptance Criteria for MDEFW Pumps Suction
Lines, Rev. 1
OX001K87, Unit 1 Support Restraint Number 1-01A-441-H4130, Rev. 2
Calculations
DPC-1381.05-00-0009, Qualified Life of Agastat E7000 Series Timing Relays, Rev. 3
KC-2079, Governor Oil Pressure Tank, Oil and Air Minimum Design Pressure, Rev. 4
KC-Unit 1-2-0098, Keowee Governor Mechanical Single Failure Analysis, Rev. 5
OSC-0864, ONS Units 1/2/3 RC System Decay Heat Removal Following Loss of Intake
Canal/Structure, Rev. 3
OSC-10180, DRIFT ANALYSIS FOR RC WIDE RANGE PRESSURE (TS SR 3.3.8.3), Rev. 0
OSC-10866, Design Basis Operating Conditions for Turbine Bypass Valves, Rev. 0
OSC-2061, U3 AC Power System Voltage & Fault Duty Analysis, Rev. 21
OSC-2152, Atmospheric Dump Valve Cv, 12/3/1987
OSC-2155, Oconee Units 1/2/3 Motor-Driven and Turbine-Driven EFW Pump NPSHa from the
Upper Surge Tank, Rev. 10
OSC-2515, Verification of Emergency Feedwater System Flow Utilizing MFW System Bypass,
Rev. 22
OSC-2820 Emergency Procedure Setpoints Calculation, Rev. 35
OSC-3198, Verification of Turbine-Driven EFW Pump Operability with Low Turbine Steam Inlet
Pressure, Rev. 0
6
OSC-3868, Overpressurization of TDEFW Pump, Valves, and Piping (PIR 3-089-0196) (TYPE
IV), Rev. 1
OSC-4276, Oconee 125Vdc Vital Instrumentation and Control Voltage Adequacy, Rev. 1
OSC-4281, System Condition Review for LPI Valves, Rev. 17
OSC-4300, Protective Relay Settings, Rev. 19
OSC-4494, GL Review for CCW Valves, Rev. 6
OSC-4701, Operability Evaluation for PIR 0-092-0057, Rev. 1
OSC-4775, Justification for Elevated UST Temperatures at 30% FP, Rev. 2
OSC-4989, Auxiliary Service Water System Hydraulic Model, Rev. 11
OSC-5093, SSF Voltage and SC Study, Rev. 14
OSC-5125, ONS Units 1/2/3 ASW NPSH Analysis, Rev. 6
OSC-5244, ONS 1, 2, 3 TDEFW Pump NPSH Analysis With Suction on the Hotwell, Rev. 2
OSC-5296, Turbine Bypass Valve Replacement, Rev. 3
OSC-5599, GL 89-10 MOV Calculation for Unit 3 Gate and Globe Valves at Oconee, Rev. 28
OSC-5599, GL89-10 Gate MOV Calculation, Rev. 28
OSC-5674, GL 89-10 MOV Calculation for Unit 1 Gate and Globe Valves at Oconee, Rev. 31
OSC-5675, GL 89-10 MOV Calculation for Unit 2 Gate and Globe Valves at Oconee, Rev. 32
OSC-5883, GL 89-10 MOV Calculation for Unit 3 Butterfly Valves at Oconee, Rev. 2
OSC-5890, Weak Link Analysis for 1, 2, 3, CCW-10, 11, 12 and 13, Rev. 0
OSC-5964, ONS Units 1-3 EFW Combined Inventory (EDFW), Rev. 8
OSC-5993, Temperature Evaluation of Plant Systems/Components During A Loss of Lake
Keowee, Rev. 3
OSC-6039, GL 95-07 Pressure Locking and Thermal Binding of LPI Valves, Rev. 10
OSC-6118, Loss of Offsite Power: Event Mitigation Requirements, Rev. 14
OSC-6144, MCC Contactor Voltage Adequacy Verification, Rev. 12
OSC-6196, Capability of the Atmospheric Dump Valve System Valves to Function During a Safe
Shutdown Facility Event, Rev. 3
OSC-6544, Determine if Main Steam Pressure Can be Controlled Using the ADV Flow Path
during an Accident that Requires Operation of the SSF ASW System, Rev. 2
OSC-7175, Oconee Nuclear Station JOG Category 1 and 2 AOV Valve List, Rev. 6
OSC-7212, Unit 1 MDEFW Pump NPSHa From the Hotwell, Rev. 6
OSC-7372, Determination of Maximum Emergency Feedwater Flow to a Depressurized Steam
Generator, Rev. 6
OSC-7608, AC Power System ETAP Model Base File, Rev. 12
OSC-7629, Nitrogen Bottle Sizing Calculation for MS-87, MS-126, and MS-129, Rev. 0
OSC-7633, Allowable Leakage Rate of Nitrogen Manifold for MS-93, Rev. 3
OSC-7745, System Review for Air-Operated Valves in the Main Steam System, Rev. 0
OSC-7775, Sizing Evaluation of Air Operated Valves MS0087, Rev. 0
OSC-7776, Sizing Evaluation of Air Operated Valve MS0093, Rev. 2
OSC-8089.01, HIGH ENERGY LINE BREAK (HELB) SAFE SHUTDOWN TARGET LIST
(SSTL), Rev. 5
OSC-8100, EFW Requirements During Unit Cooldown, Rev. 1
OSC-8181, ROTSG Tornado Protection Analysis, Rev. 4
OSC-8483, Degraded Grid Voltage Alarm Setpoints for Real Time Contingency, Rev. 19
OSC-8556, HIGH ENERGY LINE BREAK SAFE SHUTDOWN COMPONENTS ANALYSIS,
Rev. 1
OSC-8916, ADV Throttle Valve Minimum Flow Requirement Evaluation, 5/25/2006
OSC-9605, Acceptable TDEFW Recirculation Flow Rate, Rev. 0
OSC-9831, Protective Relay Settings Associated with PSW Switchgear, Rev. 4
OSC-9832, PSW AC Power System ETAP Model, Rev. 2
7
Seismic/Weak-Link Report for Valve Item Number DVM-1296, 4 Crane Gate Valve
w/Limitorque Motor Actuator, 10/30/2002
Design Basis Documents
10CFR50.59 Evaluation NSM ON-52850/0 230kV Switchyard Low Voltage Logic, 4/30/1990
ATWS Mitigation System Actuation Circuitry (AMSAC) and the Diverse Scram System (DSS),
OSS-0254.00-00-2001, Rev. 12
ENGINEERING SUPPORT DOCUMENT, Anticipated Transient Without Scram (ATWS), Rev. 9
LER 269/90-04, Undervoltage in the 230kV Switchyard, 4/30/1990
LER 269/90-05, Design Deficiency during Degraded 230kV Switchyard, 5/24/1990
NRC Final Report - Emergency Power System, 01/19/1999
NRC Letter to ONS regarding NRC Letter June 3, 1977, 12/20/1977
NRC SER for Switchyard Degraded Grid Protection, 11/14/1990
ONS Degraded Voltage LAR Submittal - New TS Sections, 10/7/1977
ONS Letter to NRC - Preliminary Switchyard Degraded Voltage Mods, 5/8/1990
ONS Letter to NRC - Proposed Switchyard Degraded Voltage Mods, 6/18/1990
ONS response to NRC June 3, 1977 Letter on Degraded Voltage, 07/21/1977
OSS-0254.00-00-1000, Emergency Feedwater and the Auxiliary Service Water Systems,
Rev. 51
OSS-0254.00-00-1003, Design Basis Specification for the Condenser Circulating Water System,
Rev. 36
OSS-0254.00-00-1028, Design Basis Specification for the Low Pressure Injection and Core
Flood System, Rev. 42
OSS-0254.00-00-1037, Design Basis Specification for the Main Steam System, Rev. 42
OSS-0254.00-00-1037, Main Steam System, Rev. 42
OSS-0254.00-00-1045, Keowee Governor Oil (OG) System, Rev. 11
OSS-0254.00-00-2000, Design Basis Spec - 4kV Essential APS, Rev. 20
OSS-0254.00-00-2004, Design Basis Spec - 230kV Switchyard System, Rev. 11
Problem Identification Process (PIP) Reports
G-10-00443
G-12-00037
G-12-00781
G-12-00781
G-12-01234
G-12-01432
G-13-00315
G-13-01455
M-09-06278
O-02-01066
O-04-01588
O-05-01401
O-05-03770
O-06-07655
O-07-01445
O-07-04681
O-07-05529
O-08-00762
O-08-05120
O-10-00664
O-10-01360
O-10-02413
O-10-11024
O-11-02208
O-11-03046
O-11-05922
O-11-06189
O-11-06189
O-11-06432
O-11-06587
O-11-06968
O-11-08439
O-11-08724
O-11-08793
O-11-08958
O-11-09142
O-11-09351
O-11-10881
O-11-10907
O-11-10954
O-11-11438
O-11-11440
O-11-11449
O-11-11453
O-11-12834
O-12-00240
O-12-00710
O-12-03322
O-12-07129
O-12-09907
O-12-10907
O-12-11006
O-12-14933
O-13-02449
O-13-03398
O-13-04629
O-13-04740
O-13-04928
O-13-05046
O-13-05132
O-13-08994
O-13-09151
O-13-09152
O-13-09586
O-13-10153
O-13-11180
O-13-11184
O-13-12270
O-13-12786
O-13-13636
O-13-13745
O-13-14542
O-14-00215
O-14-00310
O-14-01564
O-14-01592
O-14-01790
O-14-02724
8
Completed Tests
IP/0/A/3001/011K (1CCW-12), Testing MOVs Using VIPER, performed 2/5/2013
IP/0/A/3001/011K (1LP-15), Testing MOVs Using VIPER, performed 5/18/2010
IP/0/A/3001/011K (1LP-16), Testing MOVs Using VIPER, performed 5/17/2010
IP/0/A/3001/011K (2CCW-10), Testing MOVs Using VIPER, performed 4/24/2012
IP/0/A/3001/011K (3CCW-12), Testing MOVs Using VIPER, performed 11/21/2011
IP/0/A/3001/011K (3LP-15), Testing MOVs Using VIPER, performed 3/27/2013
IP/0/A/3001/011K (3LP-16), Testing MOVs Using VIPER, performed 4/21/2012
IP/0/B/0270/001S, Turbine Bypass Valves Instrument Calibration, performed 11/11/2013
Procedure PT/0/A/0251/010, Auxiliary Service Water Pump Test, Rev. 60 completed 10/8/2013
Procedure PT/0/A/0251/010, Auxiliary Service Water Pump Test, Rev. 60 completed 1/23/2014
Procedure PT/1/A/0600/012, Turbine Driven Emergency Feedwater Pump Test, Rev. 95,
completed 5/20/2011
Procedure PT/1/A/0600/012, Turbine Driven Emergency Feedwater Pump Test, Rev. 95,
completed 5/25/2011
Procedure PT/1/A/0600/012, Turbine Driven Emergency Feedwater Pump Test, Rev. 97,
completed 12/7/2012
Procedure PT/1/A/0600/013, Motor Driven Emergency Feedwater Pump Test, Rev. 71
completed 11/4/2013
Procedure PT/1/A/0600/014, Emergency Feedwater Pump Suction From Hotwell Test, Rev. 37
completed 11/18/2012
Procedure PT/2/ A/0251/014, Feedwater Check Valve Functional Test, Rev. 10, completed
11/21/2013
Procedure PT/2/A/0600/009, TD EFDWP Overspeed Test, Rev. 006, completed 11/16/2013
Procedure PT/2/A/0600/012, Turbine Driven Emergency Feedwater Pump Test, Rev. 91,
completed 11/18/2013
Procedure PT/2/A/0600/028, 2MS-93 Nitrogen Supply Leakage Test, Rev. 9, completed 5/2/13
Procedure PT/2/A/0600/028, 2MS-93 Nitrogen Supply Leakage Test, Rev. 8, completed 6/23/12
Procedure PT/3/A/0152/015, Main Steam System Valve Stroke Test, Rev. 17, completed
4/17/13
Procedure PT/3/A/0600/013, Motor Driven Emergency Feedwater Pump Test, Rev. 61,
completed 8/20/12
Procedure TT/2/A/0600/023, Turbine Driven Emergency Feedwater Pump Test, Rev. 0,
completed 7/20/02
PT/1/A/0150/054, 1LP-15 and 1LP-16 Leak Test, performed 4/7/2011, 11/3/2012
PT/1/A/0152/006, CCW System Valve Stroke Test, performed 2/10/2011, 4/5/2012, 12/9/2013
PT/1/A/0152/012, LPI System Valve Stroke Test, performed 1/27/2013, 7/14/2013, 10/6/2013,
12/29/2013
PT/1/A/0152/015, MS System Valve Stroke Test, performed 5/13/2011, 11/19/2012, 11/25/2013
PT/1/A/0251/019, MS System Atmospheric Dump Valve Functional Test, performed 11/28/2012
PT/1/A/0261/007, Dam Failure Test, performed 11/15/2012
PT/2/A/0150/054, 2LP-15 and 2LP-16 Leak Test, performed 10/19/2013
PT/2/A/0150/054, 2LP-15 and 2LP-16 Leak Test, performed 10/28/2011, 10/19/2013
PT/2/A/0152/006, CCW System Valve Stroke Test, performed 1/7/2011, 6/30/2012, 8/25/2013
PT/2/A/0152/012, LPI System Valve Stroke Test, performed 11/6/2011, 4/14/2012, 8/15/2012,
11/23/2013
PT/2/A/0152/015, MS System Valve Stroke Test, performed 11/14/2011, 11/18/2013
PT/2/A/0251/019, MS System Atmospheric Dump Valve Functional Test, performed 12/4/2013
PT/2/A/0261/007, Dam Failure Test, performed 11/9/2013
PT/3/ A/0251/014, Feedwater Check Valve Functional Test, Rev. 10, completed 5/27/2012
PT/3/A/0150/054, 3LP-15 and 3LP-16 Leak Test, performed 10/28/2010, 4/22/2012
9
PT/3/A/0152/006, CCW System Valve Stroke Test, performed 7/23/2011, 9/15/2012, 3/2/2013
PT/3/A/0152/012, LPI System Valve Stroke Test, performed 6/3/2012, 1/6/2013, 6/23/2013,
9/15/2013
PT/3/A/0152/015, MS System Valve Stroke Test, performed 5/29/2012
PT/3/A/0251/019, MS System Atmospheric Dump Valve Functional Test, performed 6/6/2012
PT/3/A/0261/007, Dam Failure Test, performed 5/16/2012
Work Orders
Miscellaneous Documents
AD-PI-ALL-0300, CDBI Readiness Self-Assessment, Rev. 0
Agastat E7000 Series Time Delay Relay Vendor Manual, March 2013
Condenser Circulating Water System Health Report, 2013 Q2, Q3; 2011 Q3, Q4
DPS-1205.19-00-0001, Limitorque Specifications, Rev. 4, 10/9/2006
EPRI Application Guide for MOVs in NPPs, Rev. 2
EPRI Technical Repair Guidelines for Limitorque Model SMB-000 Valve Actuator, Rev. 1,
Section 14.12, Actuator Hammering of Valve Sets
EQMM-1393.01-P01-04, Differential Pressure Electronic Transmitter, Rev. 4
GL 2006-02 RAI, ONS Response to RAI on GL 2006-02, 01/31/2007
GL 2006-02, ONS Response to GL 2006-02, 03/30/2006
Governor Oil Analysis Report WO No. 02096663, 10/02/2013
High Pressure Injection System Health Report, 2013 Q2, Q3; 2011 Q3, Q4
IP/0/A/2001/003A, Inspection and Maintenance of 4.16kV and 6.9kV ACB, Rev. 53
IP/0/A/3011/013, Molded Case Circuit Breaker Test and Inspection, Rev. 28
IST Basis Document for LP15/16, CCW10/11/12/13, MS19/22/28/31, and MS162/164,
February 2014
Job Performance Measure CRO-027A, Align ECCS Suction from Emergency Sump, Rev. 6
Job Performance Measure CRO-028, Align HPI/LPI Piggyback Mode, Rev. 14
Job Performance Measure CRO-047, Activate the SSF, Rev. 19
Keowee DC Power and Supersystem System Health Reports for: 1st Quarter 2011, 2nd
Quarter 2012, 3rd Quarter 2013, and 4th Quarter 2013
Low Pressure Injection System Health Report, 2013 Q2, Q3; 2011 Q3, Q4
Main Steam System Health Report, 2013 Q2, Q3; 2011 Q3, Q4
Maintenance Rule Scoping Document for the ASW system
Maintenance Rule Scoping Document for the EFW system
Maintenance Rule Scoping Document for the MS system
Maintenance Rule: Summary Sheets, CCW System, MS System, HPI System, February 2014
Motor-Operated Valve Health Report, 2013 Q2, Q3, 2012 Q1, Q2
MR Scoping Documents for ATWS System, 2/27/2014
N/A, 3rd Qtr 2013 System Health Report - SSF, 01/28/2014
N/A, 4th Qtr 2013 System Health Report - 4160V Emergency Power System, 01/27/2014
N/A, 4th Qtr 2013 System Health Report - PSW, 02/03/2014
N/A, Fleet Template - Protective Relays - Electromechanical, 12/18/2008
NSD 408, Testing, Rev. 17
10
Nuclear Scaffold Manual, Volume 1, Rev. 4
NUREG-1430, Standard Technical Specs - Babcock & Wilcox Plants Vols1&2, Rev. 4
Oconee Nuclear Station EOP/TBD Revision 10 Deviation Document, Rev. 1
OM 200B-0006-001, EFW Pump (Turbine) Instruction Book, Rev. D15
OM 206-0033 001, Motor Driven Emergency Feedwater Pumps, Re. D04
OM 206-0034 001, Motor Driven Emergency Feedwater Pumps, Re. D04
OM 208-0046 001, Auxiliary Service Water Pump Distribution Manual, Rev. D2
OM 248-0351.001, Valve Operation and Maintenance Instructions for DMV-239, 6/3/1986
OM 254-0171.001, Crosby Style HA Self Actuated Nozzle Type Safety Valves, 8/27/2007
OM 254-0409.001, Main Steam Safety Valve Cycle Life Test Program
(Test Report 5485), 3/5/2007
OM 302-0705.001, ABB Protective Relays, 11/6/2013
OM 302A-0083.001, Instruction Manual for PSW 10MVA Transformers, Rev. D7
OM 308-0356001, Vendor Instruction Book - MCC 3XSF Breaker 3A, 03/22/1995
OM 314-0279, Vendor Instruction Book - Service Water Pump Motor, 09/26/1983
OM-201.L-0037-001, Technical Manual, Model 753 Gage Pressure Electronic Transmitter,
Manual No. 85G4, 11/08/1985
OM-206.A-0005.001, Instruction Manual Turbine Driven EFW Pump, Rev. D8
OM-251.-0782.003, Instruction Manual for Type 657NS Diaphragm Actuator, Rev. 0
OM-251-0771-001, Operation and Maintenance Instructions Drag Velocity Control Element,
Rev. A
ONS Degraded Equipment List dated 11/11/2013
ONS Margin Issue List dated 11/13/2014
ONS Units 1, 2 & 3 Environmental Qualification Criteria Manual, Rev. 21, 1/30/2014
OP-OC-PNS-LPI, Low Pressure Injection System, Rev. 26g
OP-OC-SAE-R229, CPE AP EOP Exercise Guide, Rev. 1
OP-OC-SAE-S039, Exercise Guide, Rev. 00a
PMID 00039683 for MS-87
SSC Declared MR (a)(1) List 10/1/11 to 11/4/13
Station Health Group Level Report for EFW dated 7/1 to 9/30/2013
Station Health Group Level Report for Main Steam 7/1 to 9/30/2012
System Health Report, Unit 1 ATWS Health Report, 4th Qtr 2013
System Health Report, Unit 2 ATWS Health Report, 4th Qtr 2013
System Health Report, Unit 3 ATWS Health Report, 4th Qtr 2013
Corrective Action Documents Written Due to this Inspection
G-14-00547
G-14-00631
O-14-01338
O-14-01409
O-14-01416
O-14-01441
O-14-01460
O-14-01465
O-14-01531
O-14-01546
O-14-01556
O-14-01564
O-14-01578
O-14-01584
O-14-01585
O-14-01587
O-14-01588
O-14-01592
O-14-01593
O-14-01790
O-14-01792
O-14-01801
O-14-01818
O-14-01831
O-14-01910
O-14-02034
O-14-02035
O-14-02205
O-14-02215
O-14-02245
O-14-02251
O-14-02280
O-14-02282
O-14-02311
O-14-02333
O-14-02335
O-14-02337
O-14-02338
O-14-02351
O-14-02355
O-14-02361
O-14-02376
O-14-02387
O-14-02401
O-14-02416
O-14-02458
O-14-02470
O-14-02656
O-14-02724
O-14-02725
O-14-02781
O-14-02811
O-14-02837
O-14-02889
O-14-02914
O-14-02916
11
O-14-02917
O-14-02947
O-14-02953
O-14-02956
O-14-02963
O-14-02965
O-14-02965
O-14-03033
O-14-03072
O-14-03190
O-14-03915
O-14-04683
O-14-05125