ML17363A435
| ML17363A435 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 12/29/2017 |
| From: | Jeffers M NRC/RGN-III/DRS/EB2 |
| To: | Arnone C Entergy Nuclear Operations |
| References | |
| IR 2017007 | |
| Download: ML17363A435 (26) | |
See also: IR 05000255/2017007
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION III
2443 WARRENVILLE RD. SUITE 210
LISLE, ILLINOIS 60532-4352
December 29, 2017
Mr. Charles Arnone
Vice President, Operations
Entergy Nuclear Operations, Inc.
Palisades Nuclear Plant
27780 Blue Star Memorial Highway
Covert, MI 49043-9530
SUBJECT: PALISADES NUCLEAR PLANTNRC DESIGN BASES ASSURANCE
INSPECTION (TEAMS) INSPECTION REPORT 05000255/2017007
Dear Mr. Arnone:
On November 17, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed a
Triennial Baseline Design Bases Assurance Inspection (Teams) at your Palisades Nuclear
Plant. The enclosed report documents the results of this inspection, which were discussed
on November 17, 2017, with Mr. D. Corbin, and other members of your staff.
Two NRC-identified findings of very-low safety significance were identified and one Unresolved
Issue. The findings involved violations of NRC requirements. However, because of their very-low
safety significance, and because the issues were entered into your Corrective Action Program,
the NRC is treating the issues as Non-Cited Violations in accordance with Section 2.3.2 of the
If you contest the violations or significance of these Non-Cited Violations, 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 III; the Director, Office of Enforcement;
and the NRC resident inspector at the Palisades Nuclear Plant.
This letter, its enclosure, and your response (if any) will be made available for public inspection
and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room
in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely,
/RA/
Mark Jeffers, Chief
Engineering Branch 2
Division of Reactor Safety
Docket No. 50-255
License No. DPR-20
Enclosure:
cc: Distribution via LISTSERV
C. Arnone -2-
Letter from Charles Arnone to Mark Jeffers dated December 29, 2017
SUBJECT: PALISADES NUCLEAR PLANTNRC DESIGN BASES ASSURANCE
INSPECTION (TEAMS) INSPECTION REPORT 05000255/2017007
DISTRIBUTION:
RidsNrrPMPalisades Resource
RidsNrrDorlLpl3
RidsNrrDirsIrib Resource
Cynthia Pederson
Kenneth OBrien
DRPIII
DRSIII
ROPreports.Resource@nrc.gov
ADAMS Accession Number ML17363A435
OFFICE RIII RIII RIII RIII
NAME MJeffers for MJeffers
JBenjamin:cl
DATE 12/29/17 12/29/17
OFFICIAL RECORD COPY
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No: 50-255
License No: DPR-20
Report No: 05000255/2017007
Licensee: Entergy Nuclear Operations, Inc.
Facility: Palisades Nuclear Plant
Location: Covert, MI
Dates: October 30-November 17, 2017
Inspectors: J. Benjamin, Senior Reactor Inspector, Lead
B. Jose, Senior Reactor Inspector, Electrical
J. Bozga, Senior Reactor Inspector, Structural
J. Robbins, Reactor Inspector, Operations
G. Nicely, Electrical Contractor
J. Zudan, Mechanical Contractor
Approved by: Mark Jeffers, Chief
Engineering Branch 2
Division of Reactor Safety
Enclosure
SUMMARY
Inspection Report 05000255/2017007, 10/30/2017-11/17/2017; Palisades Nuclear Plant;
Design Bases Assurance Inspection (Teams).
The inspection was a 2-week onsite baseline inspection that focused on the design of
components and modifications to mitigating systems. The inspection was conducted by
regional engineering inspectors and two consultants. The inspection team identified two
findings of very-low safety significance (Green) associated with Non-Cited Violation (NCVs) of
U.S. Nuclear Regulatory Commission (NRC) regulations. The significance of inspection findings
is indicated by their color (i.e., Green, White, Yellow, and Red) and determined using Inspection
Manual Chapter 0609, Significance Determination Process, dated April 29, 2015.
Cross-cutting aspects are determined using Inspection Manual Chapter 0310, Aspects Within
the Cross-Cutting Areas, dated December 4, 2014. All violations of NRC requirements are
dispositioned in accordance with the NRCs Enforcement Policy, dated November 1, 2016. The
NRC's program for overseeing the safe operation of commercial nuclear power reactors is
described in NUREG-1649, Reactor Oversight Process, Revision 6, dated July 2016.
NRC-Identified and Self-Revealed Findings
Cornerstone: Mitigating Systems
Green. The team identified a finding of very-low safety significance (Green) and an
associated NCV of Title 10 of the Code of Federal Regulations, Part 50, Appendix B,
Criterion XI, Test Control, for the failure to periodically test the emergency diesel
generators (EDGs) capability to start and accelerate all of the sequenced loads within
the applicable design voltage and frequency transient and recovery limits. Specifically,
EDG testing activities did not demonstrate that all of the EDG auto-sequenced loads
started and accelerated within the applicable voltage and frequency limits during start-up
and recovery. In addition, the licensee did not perform adequate post-modification
testing after replacing the EDG governor controller system or voltage regulators.
The licensee captured these issues in their Corrective Action Program as Condition
Report (CR) 2017-05265 and CR 2017-05283, and performed an operability evaluation
which reasonably determined the affected structures, systems, and components were
operable.
The performance deficiency was determined to be more-than-minor because it was
associated with the Mitigating Systems cornerstone attribute of equipment performance
and affected the cornerstone objective of ensuring the availability, reliability, and
capability of mitigating systems that respond to initiating events to prevent undesirable
consequences. The finding screened as of very-low safety significance (Green)
because it did not result in the loss of operability or functionality of mitigating systems.
Specifically, the licensee evaluated the most recent voltage and frequency data from the
last EDG output breaker tests in which the data recorder was left running after the output
breaker shut and reasonably determined that the EDGs and the affected loads were
operable. The team did not identify a cross-cutting aspect associated with this finding
because it was not confirmed to reflect current performance due to the age of the
performance deficiency. Specifically, the associated testing procedures were
established more than 3 years ago. (Section 1R21.3.b(1))
2
Cornerstone: Barrier Integrity
Green. The inspectors identified a finding of low safety significance (Green) and an
associated potential NCV of Title 10 of the Code of Federal Regulations, Part 50,
Appendix B, Criterion III, Design Control, for failure to meet Updated Final Safety
Analysis Report requirements for containment spray piping supports, specifically straps.
Specifically, the inspectors identified that Calculation No. EA-SP-03369-02, Revision 0,
used inelastic acceptance limits for the pipe straps which connect the pipe to the pipe
support, in order to demonstrate Class I compliance which was not in accordance with
the design and licensing basis specification. The license entered the issue into their
Corrective Action Program as CR-PLP-2017-05246, Spray Pipe Support, dated
November 14, 2017. The licensee performed an analysis to establish reasonable
assurance of operability and the inspectors with support from the Office from the Nuclear
Reactor Regulation reviewed this operability and no performance deficiencies were
identified.
The performance deficiency was determined to be more-than-minor because it was
associated with the Barrier Integrity Cornerstone attribute of design control and
adversely affected the cornerstone objective to provide reasonable assurance that
physical design barriers (fuel cladding, reactor coolant system, and containment) protect
the public from radionuclide releases caused by accidents or events. This finding is of
very-low safety significance (Green) because there was no actual reactor containment
barrier degradation. The inspectors did not identify a cross-cutting aspect associated
with this finding because this was a legacy design issue; and therefore, was not
reflective of current performance. (Section 1R21.5.b(1))
3
REPORT DETAILS
1. REACTOR SAFETY
Cornerstone: Initiating Events, Mitigating Systems, and Barrier Integrity
1R21 Design Bases Assurance Inspection (Teams) (71111.21M)
.1 Introduction
The objective of the design bases assurance inspection is to verify that design bases
have been correctly implemented for the selected risk significant components and
modifications, and that operating procedures and operator actions are consistent with
design and licensing bases. As plants age, their design bases may be difficult to
determine and an important design feature may be altered or disabled during a
modification. The inspection also monitors the implementation of modifications to
structures, systems, and components as modifications to one system may also affect the
design bases and functioning of interfacing systems as well as introduce the potential for
common cause failures. The Probabilistic Risk Assessment model assumes the
capability of safety systems and components to perform their intended safety function
successfully. This inspectable area verifies aspects of the Initiating Events, Mitigating
Systems, and Barrier Integrity cornerstones for which there are no indicators to measure
performance.
Specific documents reviewed during the inspection are listed in the Attachment to the
report.
.2 Inspection Sample Selection Process
The team initially selected samples that were more safety-significant for shutdown
conditions based upon the companys public announcement to permanently shutdown in
the fall of 2018. Components related to the spent fuel pool cooling water system, spent
fuel pool reactivity control components, and control room emergency ventilation systems
were therefore initially chosen.
The licensee made the decision to not shutdown in 2018, but rather continue operations
until the spring of 2022 several weeks before the inspection was scheduled to start.
Based upon this information, the team selected additional, more traditional, risk
significant components based upon the risk insights of the licensees Probabilistic Risk
Assessment model, and the Palisades Nuclear Standardized Plant Analysis Risk model
with the assistance of a U.S. Nuclear Regulatory Commission (NRC) Region III senior
risk analyst. The team selected both the loss of off-site power (LOOP) and steam
generator tube rupture events to refine the final component selection, including the large
early release frequency (LERF) component selection.
The team also used additional component information such as a margin assessment in
the selection process. This design margin assessment considered original design
reductions caused by design modifications, power uprates, or reductions due to
degraded material condition. Equipment reliability issues were also considered in the
selection of components for detailed review. These included items such as performance
test results, significant corrective actions, repeated maintenance activities, Maintenance
Rule (a)(1) status, components requiring an operability evaluation, system health
4
reports, and NRC resident inspector input of problem areas/equipment. Consideration
was also given to the uniqueness and complexity of the design, operating experience,
and the available defense in depth margins. A summary of the reviews performed and
the specific inspection findings identified are included in the following sections of the
report.
The team also identified modifications for review with a focus on modifications
implemented within the last 3 years. In addition, the inspectors selected procedures and
operating experience issues associated with the selected components and other risk
informed factors.
This inspection constituted 21 samples (10 components with 1 component associated
with LERF implications, 6 modifications, and 5 operating experience) as defined in
Inspection Procedure 71111.21M-02.01. The team applied approximately 80 percent
inspection effort on the traditional risk-significant component, operating experience, and
modification samples. The remaining approximate 20 percent effort was used to review
the original shutdown component selection.
.3 Component Design
a. Inspection Scope
The team reviewed the Updated Final Safety Analysis Report (UFSAR), Technical
Specifications, design basis documents, drawings, calculations and other available
design basis information, to determine the performance requirements of the selected
components. The team used applicable industry standards, such as the American
Society of Mechanical Engineers Code, Institute of Electrical and Electronics
Engineers (IEEE) Standards, and the National Electric Code, to evaluate acceptability
of the systems design. The NRC also evaluated licensee actions, if any, taken in
response to NRC issued operating experience, such as Bulletins, Generic Letters,
Regulatory Issue Summaries, and Information Notices. The review was to verify that the
selected components would function as designed when required and support proper
operation of the associated systems. The attributes that were needed for a component
to perform its required function included process medium, energy sources, control
systems, operator actions, and heat removal. The attributes to verify that the component
condition and tested capability was consistent with the design bases and was
appropriate may have included installed configuration, system operation, detailed
design, system testing, equipment and environmental qualification, equipment
protection, component inputs and outputs, operating experience, and component
degradation.
For each of the components selected, the inspectors reviewed the maintenance history,
preventive maintenance activities, system health reports, operating experience-related
information, vendor manuals, electrical and mechanical drawings, operating procedures,
and licensees Corrective Action Program (CAP) documents. Field walkdowns were
conducted for all accessible components selected to assess material condition, including
age-related degradation, configuration, potential vulnerability to hazards, and
consistency between the as-built condition and the design. In addition, the team
interviewed licensee personnel from multiple disciplines such as operations,
engineering, and maintenance. Other attributes reviewed are included as part of the
scope for each individual component.
5
The following 10 components (samples), including a component with LERF implications,
were reviewed:
Emergency Diesel Generator (EDG) Fuel Oil Transfer Pump (MDP 18A): The
team reviewed drawings and calculations associated with pump motor cable
sizing, voltage drop during degraded voltage conditions, and total current drawn
by the motor compared to the ampacity of the feeder cables. The team also,
reviewed the maximum short circuit current available at the feeder breaker and
verified that the breaker interrupting capacity was not exceeded. The team also,
reviewed the vendor manual of the pump motor to compare recommended
maintenance versus the licensees actual maintenance activities. The team
reviewed calculations for hydraulic performance, net positive suction
head (NPSH), required total developed head, pump vibration analysis, pump
run-out conditions, potential for vortex formation and loss of suction at the suction
source. Seismic design documentation was reviewed to verify pump design was
consistent with limiting seismic design conditions. The team also reviewed diesel
fuel oil storage requirements for transient and accident conditions.
Condensate Storage Tank (CST): The team reviewed seismic and tornado
missile calculations, drawings, and operating procedures associated with the
CST. The inspectors assessed the stress analysis for portions of the auxiliary
feedwater piping that is connected to the CST. The inspectors assessed the
tank's volume, capacity, levels, and setpoints with respect to auxiliary
feedwater (AFW) pump suction requirements.
CST Level Transmitters (LT021/22): The team reviewed the schematic and
instrument loop diagrams of the level transmitters, their setting sheets, power
supply requirements, calibration data and heat trace requirements of the
instrument tubing. The team reviewed the operation of the level transmitters
during freezing conditions concurrent with a loss of offsite power or a station
black out, since the heat tracings were powered from nonsafety-related sources.
The team reviewed the trip circuitry of the AFW pumps on low suction pressure
and verified that the low suction pressure switch settings were adequately
coordinated with the AFW pump NPSH requirements.
Positive Displacement Pump (MDP 55A): The team reviewed the electrical
schematic and wiring drawings of the pump motor, name plate data, and
minimum voltage requirements compared to the minimum available voltage
during degraded voltage conditions. The team also reviewed the feeder cable
size compared to the pump motor ampacity requirements. The team verified that
the motor name plate data was correctly translated in the electrical load flow,
short circuit and voltage drop calculations. Also, the feeder breaker interrupting
capacity was verified to be above the maximum short circuit current available at
the breaker and the breaker control components. The team reviewed
calculations for hydraulic performance, NPSH, required total developed head and
pump run-out conditions. Seismic design documentation was reviewed to verify
that the pump design was consistent with limiting seismic design conditions.
Test results were reviewed to verify acceptance criteria were met and
performance degradation would be identified, taking into account set-point
tolerances and instrument inaccuracies.
2400 VAC Safeguard Transformer: The team reviewed loading calculations to
determine whether the capacity of the transformer was adequate to supply
worst-case loading. Voltage calculations and operating procedures were
reviewed to determine whether transformer taps and administrative controls for
6
switchyard voltage were adequate to assure the availability of offsite power
during accident conditions. Procedures for preventive maintenance, inspection,
and testing were reviewed to compare maintenance practices against industry
and vendor guidance.
2400 VAC Safeguard Bus: The team reviewed calculations for electrical
distribution, system load flow/voltage drop, short-circuit, and electrical protection
to verify that bus capacity and voltages remained within the minimum acceptable
limits. The protective device settings and circuit breaker ratings were reviewed to
ensure adequate selective protection coordination of connected equipment
during worst-case short circuit conditions. The team verified that degraded and
loss of voltage relays and associated time delays were set in accordance with
calculations, and that associated calibration procedures were consistent with
calculation assumptions, associated time delays, and set point accuracy
calculations. The team evaluated selected portions of the licensee response to
NRC Generic Letter 2006-02, Grid Reliability and the Impact on Plant Risk and
the Operability of Offsite Power, dated February 1, 2006. The team reviewed
the stations interface and coordination with the transmission system operator for
plant voltage requirements and notification set points. The team reviewed the
125VDC voltage drop calculations to ensure that the EDG breaker controls
would have adequate voltage to operate during a LOOP event. The EDG
loading calculations were reviewed to determine whether the capacity of the
EDGs was adequate to supply worst case accident loads. The last EDG
LOOP/loss-of-coolant accident (LOCA) surveillance tests were reviewed to
ensure that the voltage and frequency dips and recovery were within the design
limits.
Train A Main Steam Isolation Valve (CV-0510)LERF Sample: The inspectors
reviewed Inservice Testing (IST) stroke time data and Air Operated Valve Program
requirements to ensure valve performance was being appropriately monitored.
The inspectors reviewed the main steam piping from steam generator A to the
containment penetration pipe stress analysis. The team reviewed open and
closing force calculations to assess the main steam isolation valves capability to
function as described in the UFSAR under all bounding conditions. The team
reviewed power and control wiring diagrams to assess the control and actuation
schemes adequacy. This review constituted one component sample with LERF
implications.
AFW Suction Line Low Pressure Switch (PS 0741): The team reviewed the
instrument loop and power supply drawings, pressure switch set points and
calibration data. The team also, reviewed the AFW pumps low suction pressure
trip circuitry and verified that the settings were adequately coordinated to ensure
that the AFW pumps had adequate NPSH under design basis conditions.
A Train Spent Fuel Pool Cooling Water Pump: The team reviewed calculations
for required hydraulic performance, NPSH, required total developed head, pump
vibration analysis, pump run-out conditions, potential for vortex formation, loss of
suction at the suction source and alternate means of fuel pool purification with a
floating skimmer. Seismic design documentation was reviewed to verify pump
design was consistent with design based seismic conditions.
7
A Train Control Room Recirculation Ventilation Dampers: The team reviewed
calculations for load flow/voltage drop and short-circuit to verify that bus capacity
and voltages remained within minimum acceptable limits. The protective device
settings and circuit breaker ratings were reviewed to ensure adequate selective
protection coordination of connected components during worst-case short circuit
conditions.
b. Findings
(1) Failure to Periodically Test the Emergency Diesel Generators Capacity to Start and
Accelerate Design Basis Sequenced Loads
Introduction: The team identified a finding of very-low safety significance (Green)
and an associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal
Regulations (CFR), Part 50, Appendix B, Criterion XI, Test Control, for the failure
to periodically test the EDGs capability to start and accelerate all of the sequenced
loads within the applicable design voltage and frequency transient and recovery
limits. Specifically, EDG testing activities did not demonstrate that all of the EDG
auto-sequenced loads started and accelerated within the applicable voltage and
frequency limits during start-up and recovery. In addition, the licensee did not perform
adequate post-modification testing after replacing the EDG governor controller system
or voltage regulators.
Description: The UFSAR, Section 8.1.1, Design Basis, states in part that the
engineered safeguards electrical system is intended to meet all the other
requirements identified in IEEE 308-1978. The IEEE 308-1978, Section 7.4,
Periodic Equipment Tests, states that Tests shall be performed at scheduled
intervals to (1) Detect the deterioration of the system towards an unacceptable condition.
(2) Demonstrate that standby power equipment and other components that are not
exercised during normal operation of station are operable. The UFSAR, Section 8.4.1.3,
Design Basis, states that the recovery time for the EDG voltage to return to 90 percent
of rated voltage after application of each load step is less than 3 seconds. The UFSAR,
Sections 5.1.3.8 and 5.1.3.9, for Criterion 17 and 18, states that the onsite power system
can be periodically tested to assure that they are operable and functional.
The licensee is not committed to NRC Regulatory Guides (RGs) 1.108 and 1.9,
however these RGs describe an acceptable approach to test the diesels generator.
Position C.2.a.2 of RG 1.108 states that testing of diesel generator units during the Plant
Preoperational Test Program and at least once every 18 months should demonstrate
proper operation for design-accident-loading-sequence to design-load requirements and
verify that voltage and frequency are maintained within required limits. Position C.4 of
RG 1.9 stated, in part, that at no time during the loading sequence should the
frequency and voltage decrease to less than 95 percent of nominal and 75 percent of
nominal respectively. It also stated that Frequency should be restored to within
2 percent of nominal, and voltage should be restored to within 10 percent of nominal
within 60 percent of each load-sequence time interval. The licensee did not provide the
team an acceptable alternative to the requirements stated in the above RGs, but rather
do not evaluate or review the voltage and frequency responses obtained during the
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RT-8C or RT-8D surveillances. Therefore, if left uncorrected, the governor or voltage
regulator control system could, due to aging, drift from the original settings and not allow
the EDG to recover quick enough after large load sequencing and have the potential of
loads overlapping during application to the EDG.
The team noted the following deficiencies related to the EDG periodic testing:
The licensee, in surveillance procedures RT-8C and RT-8D, are only evaluating
the steady-state voltage and frequency at the EDG terminals after the load
sequencing is complete. However, section 6.3 of the test procedure verifies that
the EDG successfully supports the sequenced starting of the engineered
safeguards equipment, but only refers to Attachment 5 which only verifies the
measured times for the sequence timers.
When the electronic governor or electronic voltage regulator are replaced or
otherwise adversely affected during maintenance activities, the licensee tunes
the new device using an approximate approach in an unloading condition which
yield a gross setting.
The license entered the issue into their CAP CR 2017-05265 and CR 2017-05283. The
licensee identified that historic LOOP/LOCA frequency and voltage trace data was
available based upon the frequency and voltage recorders continuing to run during EDG
output break time testing. The licensee analyzed the last EDG LOOP/LOCA traces and
determined that the EDGs were currently operable. At the end of the inspection, the
licensee was in the process of developing the corrective action to restore compliance.
Analysis: The team determined that the failure to periodically test the EDG capacity to
start and accelerate all of the sequenced loads within the applicable voltage and
frequency limits was contrary to 10 CFR Part 50, Appendix B, Criterion XI, Test
Control, and was a performance deficiency. The performance deficiency was
determined to be more than minor because it was associated with the Mitigating
Systems cornerstone attribute of equipment performance and affected the cornerstone
objective of ensuring the availability, reliability, and capability of mitigating systems to
respond to initiating events to prevent undesirable consequences. Specifically, the
failure to test the EDGs capacity to start and accelerate all of the sequenced loads within
the applicable limits periodically to identify degradation and following maintenance
activities were the EDGs frequency and voltage responses could be impacted did not
ensure availability, reliability, and capability of components supplied by the EDGs to
perform their intended safety function.
The team determined the finding could be evaluated using the Significance
Determination Process in accordance with Inspection Manual Chapter (IMC) 0609,
Significance Determination Process, Attachment 0609.04, Initial Characterization of
Findings, issued on June 19, 2012. Because the finding impacted the Mitigating
Systems and Barrier Integrity cornerstones, the team screened the finding through
IMC 0609 Appendix A, The Significance Determination Process for Findings At-Power,
issued on June 19, 2012, using Exhibit 2, Mitigating Systems Screening Questions.
The finding screened as of very-low safety significance (Green) because it did not result
in the loss of operability or functionality of mitigating systems. Specifically, the licensee
evaluated the most recent voltage and frequency data from the last EDG output breaker
tests in which the data recorder was left running after the output breaker shut and
reasonably determined that the EDGs and the affected loads were operable.
9
The team did not identify a cross-cutting aspect associated with this finding because the
performance deficiency was not reflective of current performance due to the age of the
issue. Specifically, the associated testing procedures were established more than
3 years ago.
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion XI, Test Control, requires, in
part, that a test program be established to assure that all testing required to demonstrate
that structures, systems, and components will perform satisfactorily in service is
identified and performed in accordance with written test procedures which incorporate
the requirements and acceptance limits contained in applicable design documents. It
also stated that test results shall be documented and evaluated to assure that test
requirements have been satisfied.
The UFSAR, Section 8.1.1, Design Basis, states in part that the engineered
safeguards electrical system is intended to meet all the other requirements identified
in IEEE 308-1978. The IEEE 308-1978, Section 7.4, Periodic Equipment Tests, states
that Tests shall be performed at scheduled intervals to (1) Detect the deterioration of
the system towards an unacceptable condition. (2) Demonstrate that standby power
equipment and other components that are not exercised during normal operation of
station are operable.
Contrary to the above, as of November 15, 2017, the licensee failed to establish a
testing program to demonstrate that the EDGs could start and accelerate their
sequenced loads within the applicable voltage and frequency acceptance limits
periodically as required by IEEE 308-1978 and following maintenance activities that
could adversely affect EDG frequency and voltage response (e.g. governor and voltage
regulator maintenance activities.) The licensee is still evaluating its planned corrective
actions, however, the team determined that the continued non-compliance does not
present an immediate safety concern because the licensee reasonably determined the
affected systems, structures, and components remained operable.
Because this violation was of very-low safety significance and was entered into the
licensees CAP as CR-PLP-2017-05265 and CR-PLP-2017-05283, this violation is being
treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy.
(NCV 05000255/2017007-01; Failure to Periodically Test the Emergency Diesel
Generators Capacity to Start and Accelerate Design Basis Sequenced Loads)
.4 Mitigating System Modifications
a. Inspection Scope
The team reviewed five permanent plant modifications. This review included in-plant
walkdowns for accessible portions of the modified structures, systems, and components.
The team reviewed the modifications to verify that the design bases, licensing bases,
and performance capability of the components had not been degraded through
modifications. The modifications were selected based upon risk significance, safety
significance, and complexity. The team reviewed the modifications selected to
determine if:
the supporting design and licensing basis documentation was updated;
the changes were in accordance with the specified design requirements;
10
the procedures and training plans affected by the modification have been
adequately updated;
the test documentation as required by the applicable test programs has been
updated; and
post-modification testing adequately verified system operability and/or
functionality.
The team also used applicable industry standards to evaluate acceptability of the
modifications. The modifications listed below were reviewed as part of this inspection
effort:
Engineering Change (EC) 0000058140; Install Permanent Shielding on Letdown
Heat Exchanger;
EC 0000058141; Install Shielding on Pressurizer Surge Line E-50A Platform;
EC 0000056644; Supplemental Diesel Generator Fuel Oil Tank to Comply with
Michigan Fire Code;
EC 0000048188; FLEX EC#21 - Turbine Driven AFW System FLEX Upgrades;
EC 0000055367; Install Larger Size Power Cables between EX-04 (SU1-2) and
2400 VAC Buses 1C and 1D; and
EC 0000071766;52-389, Replace Control Transformer on CV-1510 MSIV
Bypass Valve.
b. Findings
No findings were identified.
.5 Operating Experience
a. Inspection Scope
The team reviewed five operating experience issues (samples) to ensure that generic
concerns had been adequately evaluated and addressed by the licensee. The operating
experience issues listed below were reviewed as part of this inspection:
Point Beach Containment Dome Truss License Amendment Request 278,
Risk-Informed Approach To Resolve Construction Truss Design Code
Non-Conformances; March 31, 2017;
CR-PLP-2014-04976; C and D Batteries Part 21 Separator Misalignment LCR,
KCR, and LCY Batteries;
NCV 05000285/2012011-04; Inadequate Design Basis Documentation;
NRC Information Notice 2012-06; Design Vulnerability in Electric Power
Systems; and
NRC Regulatory Information Summary 2011-012, Adequacy of Station Electrical
Distribution System Voltages.
11
b. Findings
(1) Containment Spray Pipe Support Strap Deficiencies
Introduction: The inspectors identified a finding of low safety significance (Green) and
an associated potential NCV of 10 CFR Part 50, Appendix B, Criterion III, Design
Control, for failure to meet UFSAR requirements for containment spray (CS) piping
supports, specifically straps. Specifically, the strap (connection between pipe and
support) design of CS pipe supports HC44-R884, HC44-R884.1, HC44-R884.2, HC44-
R884.3, HC44-R884.4, HC44-R884.5 and HC44-R884.9 did not comply with UFSAR
Section 5.10.1.2 and Specification No. C-173(Q) requirements.
Description: The CS system per UFSAR, Section 6.2.1, has the following safety-related
design basis functions: The function of the CS system is to limit the containment
building pressure rise and reduce the airborne radioactivity in containment by providing a
means for spraying the containment atmosphere after occurrence of a LOCA or a main
steam line break. The CS piping and pipe supports were designed to Class I
requirements as described in UFSAR, Section 5.10.1.2, titled CP Co Design Class 1
Pipe Supports and Specification No. C-173(Q), Technical Requirements for the Analysis
and Design of Safety-Related Pipe Supports, Revision 6. This specification was
classified as safety-related. Calculation No. EA-SP-03369-02, Containment Spray
System Pipe Supports, Revision 0 evaluated CS pipe supports HC44-R884,
HC44-R884.1, HC44-R884.2, HC44-R884.3, HC44-R884.4, HC44-R884.5 and
HC44-R884.9 in accordance with Class I requirements for all design basis loading.
The pipe supports were analyzed to withstand applied stress due to dead loads, live
loads, seismic loads, and thermal loads. The inspectors identified that in Calculation
No. EA-SP-03369-02, Revision 0, the licensee used inelastic acceptance limits for the
pipe straps which connect the pipe to the pipe support, in order to demonstrate Class I
compliance which was not in accordance with the design and licensing basis. The
Class I requirements were based on UFSAR, Section 5.10.1.2, and Specification
No. C-173(Q). The UFSAR, Section 5.10.1.2, does not specify the use of inelastic
capacity for the straps which are considered catalog items. The capacity is based on a
specified load capacity which is based on the strap maintaining its structural integrity
with no permanent or plastic deformation allowed when subjected to the design loading.
Specification No. C-173(Q) delineated requirements consistent with UFSAR,
Section 5.10.1.2. The inspectors determined the use of an inelastic acceptance limits
for pipe support straps did not meet Class I requirements. The license entered the
issue into their CAP as CR-PLP-2017-05246, Spray Pipe Support, dated
November 14, 2017. The licensee performed an analysis to establish reasonable
assurance of operability and the inspectors with support from the Office from the Nuclear
Reactor Regulation.
Analysis: The inspectors determined the licensees failure to meet Class I requirements
for the CS pipe supports HC44-R884, HC44-R884.1, HC44-R884.2, HC44-R884.3,
HC44-R884.4, HC44-R884.5 and HC44-R884.9 was contrary to 10 CFR Part 50,
Appendix B, Criterion III, Design Control, and was a performance deficiency. The
finding was determined to be more than minor because the finding was associated with
the Barrier Integrity Cornerstone attribute of design control and adversely affected the
cornerstone objective to provide reasonable assurance that physical design barriers (fuel
cladding, reactor coolant system, and containment) protect the public from radionuclide
releases caused by accidents or events. Specifically, failure to comply with Class I
12
requirements did not ensure the Pipe Supports HC44-R884, HC44-R884.1,
HC44-R884.2, HC44-R884.3, HC44-R884.4, HC44-R884.5 and HC44-R884.9 would
function during a Class I design basis event and would adversely affect the CS piping
system and containment barrier. The inspectors determined the finding could be
evaluated using the SDP in accordance with IMC 0609, Appendix A, The Significance
Determination Process for Findings At-Power, dated June 19, 2012, Exhibit 3,
Barrier Integrity Screening Questions, for the Barrier Integrity cornerstone (Reactor
Containment). The inspector answered no to the Barrier Integrity questions for Reactor
Containment. The finding screened as having very-low safety significance (Green).
The inspectors determined there was no cross-cutting aspect associated with this finding
because the deficiency was a legacy design calculational issue and, therefore, was not
indicative of licensees current performance.
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires,
in part, that measures be established to ensure the applicable regulatory requirements
and the design basis are correctly translated into specifications, drawings, procedures,
and instructions. The design control measures shall provide for verifying or checking the
adequacy of design.
Contrary to the above, as of November 14, 2017, the design control measures failed to
conform to Class I requirements and also failed to verify the adequacy of the design.
Specifically, Calculation No. EA-SP-03369-02, Revision 0, failed to verify the adequacy
of the design for the CS pipe supports HC44-R884, HC44-R884.1, HC44-R884.2,
HC44-R884.3, HC44-R884.4, HC44-R884.5 and HC44-R884.9 to ensure it met the
Class I requirements. Because this violation was of very-low safety significance (Green)
and it was entered into the licensees CAP as CR-PLP-2017-05246, this violation is
being treated as a NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy.
(NCV 05000255/2017007-02; Containment Spray Pipe Support Strap Deficiencies)
(2) Containment Dome Truss Analysis
Introduction: The inspectors identified an unresolved item (URI) concerning the analysis
that demonstrated the design adequacy of the containment dome truss under design
and licensing basis loading conditions.
Description: The dome truss system was originally designed to support the containment
liner plate and wet concrete during the construction of the containment dome (i.e., the
liner plate initially acted as a form and the truss supported the form). After the concrete
cured, the dome truss system was lowered away from the liner and was used to support
the safety injection tanks (SITs) and CS system piping and their associated supports.
The CS and SIT systems are both safety-related which were required to be evaluated for
seismic loads (self-weight and externally applied loads). The dome truss system would
have also been required to be evaluated for seismic loads.
The UFSAR, Section 6.1, described the safety-related design function of the SIT
system was to prevent fuel and cladding damage that could interfere with adequate
emergency core cooling, and to limit the cladding-water reaction to less than
approximately 1 percent for all break sizes in the primary system piping up to and
including the double-ended rupture of the largest primary coolant pipe, for any break
location, and for the applicable break time. Also, the SIT system also functions to
13
provide rapid injection of large quantities of borated water for added shutdown capability
during rapid cooldown of the primary system caused by a rupture of a main steam line.
UFSAR Section 6.2.1 described the safety-related design function of the CS system was
to limit the containment building pressure rise and reduce the airborne radioactivity in
containment by providing a means for spraying the containment atmosphere after
occurrence of a LOCA or a main steam line break.
The inspectors requested the design basis analysis of the dome truss system that
considers the LOCA loading on the dome truss system as well as the seismic loading
due to the applied design loads from the CS and SIT system. During the time of the
inspection, the licensee was unable to locate the dome truss analysis.
In response to the inspectors concern, the licensee entered the issue into their CAP
as CR 2017-05016, Dome Trusses, dated November 1, 2017. The licensee is
investigating the containment dome truss analysis further with the vendor of the dome
truss system.
This issue is a URI pending additional inspector review of the design basis analysis for
the containment dome truss system. (URI 05000255/2017007-03; Containment Dome
Truss Analysis)
.6 Operating Procedure Accident Scenarios
a. Inspection Scope
The team performed a detailed reviewed of the procedures listed below. The
procedures were compared to UFSAR, design assumptions, and training materials to
assess their consistency. The following operating procedures were reviewed in detail:
4.48, Time Critical Action/Time Sensitive Action Program Standard, Revision 6;
EOP TCA, EOP Time Critical/Time Sensitive Operator Action Basis, Revision 2;
SOP 22, Emergency Diesel Generators, Revision 74;
AOP 41, Alternate Safe Shutdown Procedure, Revision 3; and
AOP Supplement 8, Operation of Panels EC-150/EC-150A, Revision 0.
For the procedures listed, time dependent operator actions were reviewed for adequacy.
This review included walk downs of in-plant actions with a licensed operator. In addition,
the team evaluated operations interfaces with other departments such as engineering.
The following operator actions were reviewed:
Time critical operator actions to switch Control Room HVAC to Emergency Mode
b. Findings
No findings identified.
14
4. OTHER ACTIVITIES
4OA2 Identification and Resolution of Problems
.1 Review of Items Entered Into the Corrective Action Program
a. Inspection Scope
The team reviewed a sample of problems identified by the licensee associated with the
selected samples and that were entered into the CAP. The team reviewed these issues
to verify an appropriate threshold for identifying issues and to evaluate the effectiveness
of corrective actions related to design issues. In addition, corrective action documents
written on issues identified during the inspection were reviewed to verify adequate
problem identification and incorporation of the problem into the CAP. The specific
corrective action documents sampled and reviewed by the team are listed in the
attachment to this report.
The team also selected seven issues identified during previous component design basis
inspections to verify that the concerns were adequately evaluated and corrective actions
were identified and implemented to resolve the concern, as necessary. The following
issues were reviewed:
NCV 05000255/2014008-02, Undersized Supply Cables from Startup
Transformer to 2400V Buses;
NCV 05000255/2014008-03, Undersized Motors;
NCV 05000255/2014008-05, Lack of Analysis for Electrical Containment
Penetration Protection;
NCV 05000255/2014008-04, Failure to Ensure that 480VAC System Voltages
do not exceed Equipment Ratings;
NCV 05000255/2014008-09, Failure to Include the Degraded Voltage Channel
Time Delay in Technical Specification Surveillance Requirements;
NCV 05000255/2014008-10, Failure to include the Degraded Voltage Time
Delay in TS Surveillance Requirements; and
NCV 05000255/2014008-13, Non-conservative Surveillance for Emergency
Diesel Generator Largest Load Reject Test.
b. Findings
No findings were identified.
4OA6 Management Meetings
.1 Interim Meeting Summary
On November 16, 2017, the team presented the preliminary inspection results to
Mr. J. Hardy and other members of the licensee staff. The licensee acknowledged
the issues presented. The team confirmed that several documents reviewed were
considered proprietary and were handled in accordance with the NRC policy related to
proprietary information.
15
.2 Exit Meeting Summary
On November 17, 2017, the team presented the inspection results to Mr. D. Corbin and
other members of the licensee staff. The licensee acknowledged the issues presented.
The team asked the licensee whether any materials examined during the inspection
should be considered proprietary. Several documents reviewed by the team were
considered proprietary information and were either returned to the licensee or handled in
accordance with NRC policy on proprietary information.
ATTACHMENT: SUPPLEMENTAL INFORMATION
16
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
C. Arnone, Vice President
D. Corbon, General Manager Plant Operations
O. Gustafson, Regulatory Assurance and Performance Improvement Director
K. OConnor, Engineering Director
J. Hardy, Regulatory Assurance Director
B. Sova, Design Engineering Manager
B. Baker, Maintenance Manager
U.S. Nuclear Regulatory Commission
M. Jeffers, Branch Chief
J. Benjamin, Senior Reactor Inspector
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Close
05000255/2017007-01 NCV Failure to Periodically Test the Emergency Diesel
Generators Capacity to Start and Accelerate Design
Basis Sequenced Loads (1R21.3.b(1))05000255/2017007-02 NCV Containment Spray Pipe Support Strap Deficiencies
(1R21.5.b(1))
Open
05000255/2017007-03 URI Containment Dome Truss Analysis (1R21.5.b(2))
Discussed
None
Attachment
LIST OF DOCUMENTS REVIEWED
The following is a list of documents reviewed during the inspection. Inclusion on this list does
not imply that the NRC inspectors reviewed the documents in their entirety, but rather, that
selected sections of portions of the documents were evaluated as part of the overall inspection
effort. Inclusion of a document on this list does not imply NRC acceptance of the document or
any part of it, unless this is stated in the body of the inspection report.
50.59 SCREENINGS
- 50.59 Screening 96-2716, Replace Pumps 18A/B with New Pumps; 12/15/1996
CALCULATIONS
- EA-ELEC-EDSA-06; Palisades AC Power System - Short Circuit Analysis; Revision 2
- EA-ELEC-EDSA-10; DC System Battery D01 EDSA Model Development and Load Flow
Analysis; Revision 1
- EA-ELEC-EDSA-11; DC System Battery D02 EDSA Model Development and Load Flow
Analysis; Revision 2
- EA-ELEC-VOLT-051; MCC Power Circuit Minimum Required Voltage Analysis; Revision 2
- EA-ELEC-LDTAB-009; Battery Sizing for the Palisades Class 1E Station Batteries D01 and
D02; Revision6
- EA-SP-03374-01; Main Steam Piping Analysis Piping from Steam Generator E-50A to
Containment Penetration 2; Revision 2
- EA-POC0007899-T2; Evaluation of Condensate Storage Tank T-2; Revision 1
- EA-SP-03370; Piping Analysis for Containment Spray Piping; Revision 2
- EA-SP-03369-01; Piping Evaluation for Containment Spray Piping; Revision 2
- EA-SP-03369-02; Piping Evaluation for Containment Spray Piping; Revision 0
- EA-SP-05904-01; Pipe Stress Analysis Auxiliary Feedwater Piping; Revision 4
- EA-SP-05901-01; Pipe Stress Analysis Auxiliary Feedwater Pump Suction; Revision 0
- EA-SP-03356-01; Auxiliary Feedwater Pump Suction; Revision 0
- EA-SP-03342-02; Piping Stress Analysis for 2 Auxiliary Feedwater Pump P-8A and B
Recirculation Piping; Revision 0
- EA-SP-03342-01; Auxiliary Feedwater Discharge Piping; Revision 0
- EA-EC8083-01; Evaluation of CST for Tornado Loads; Revision 1
- EA-T-343-03; Determination of the Fuel Oil Transfer Pump Rates to the Diesel Generator Day
Tanks; 05/12/1994
- EA-C-PAL-98-1748-02; Evaluation of Allowable Leakage Rate from the Spent Fuel Pool
Cooling System; 03/15/1999
- EA-SC-96-051-01; Fuel Oil Transfer Pump Replacement; 03/06/1997
- EA-FC-958-05; Hydraulic Analysis for P-18A/B Pump Replacements; 03/06/1997
- EA-EC6432-01; Palisades Emergency Diesel Generator Diesel Fuel Oil Storage
Requirements; 05/24/2010
- EA-E-PAL-94-010-01; Alternate Diesel Generator Air Driven Diaphragm Pump Flow;
06/09/1994
- EA-EC7120-01; Auxiliary Feedwater Pumps Low Suction Pressure TripsSetpoint Change;
01/05/2009
- EA-A-PAL-94-095; Auxiliary Feedwater Pumps Net Positive Suction Head; 06/10/1994
- EA-ELEC-EDSA-03; LOCA with Offsite Power Available; Revision 2
- EA-ELEC-EDSA-04; Second Level UV Relay Setpoint Determination; Revision 0
- EA-ELEC-EDSA-06; Short Circuit Analysis; Revision 2
2
- EA-ELEC-LDTAB-005; EDG Steady State Loading Calculation; Revision 10
- EA-ELEC-VOLT-037; Degraded Voltage Calc for Safety-Related MOVs; Revision 3
- EA-ELEC-VOLT-050; MCC Control Circuit Voltage Analysis; Revision 3
- EA-ELEC-VOLT-051; MCC Power Circuit Required Voltage Analysis; Revision 2
- EA-ELEC-VOLT-052; DC Voltage Analysis; Revision 0
- 1D/202/151; Protection CalcBus 1D Incoming Breaker; Revision 1
- 1D/203/151; Protection CalcBus 1D Incoming Breaker; Revision 2
- 11-12/9B; Protection CalcLoad Center 12 Low Side; Revision 0
- 1D/201/150-151; Protection CalvStation Power; Revision 4
- EA-GL8910-01; GL 89-10 MOV Thrust Window Calculations; Revision 11
- 1/9C; High Pressure Injection MOV MO-3009; Revision 2
- 1/4C; Low Pressure Injection MOV MO-3010; Revision 2
- EA-POC0007899; Roof Drain Pipe Analysis K6AB-4; Revision 6
- EA-ELEC-AMP-030; Capability of the 2400 V Feeder Calcs to Buses 1C and 1D from Startup
Transformer 1-2; Revision 2
CORRECTIVE ACTION PROGRAM DOCUMENTS INITIATED DURING INSPECTION
- 2017-04610 - 2017-04995 - 2017-05283
- 2017-05110 - 2017-05240 - 2017-05265
- 2017-05124 - 2017-05247 - 2017-05264
- 2017-05016 - 2017-05232 - 2017-05251
- 2017-05246 - 2017-05288 - 2017-05240
- 2017-05014 - 2017-05264 - 2017-05232
- 2017-05256 - 2017-05264 - 2017-05076
- 2017-05237 - 2017-05265
- 2017-05015 - 2017-05282
CORRECTIVE ACCTIONS PROGRAM DOCUMENTS REVIEWED
- 2014-09030 - 2013-05039 - 2017-02655
- 2016-00798 - 2010-06100 - 2016-00026
- 2016-01740 - 2017-04248 - 2014-04902
- 2012-04164 - 2017-03007 - 2014-04679
- 2017-01248 - 2017-02667 - 2014-04680
- 2016-04972 - 2017-01642 - 2014-04696
- 2015-03116 - 2017-01249 - 2014-04864
- 2014-02381 - 2015-01841 - 2014-04450
- 2013-04050 - 2014-02899 - 2014-04860
- 2013-01381 - 2014-01437 - 2012-01245
- 2011-05337 - 2013-05294 - 2012-06773
- 2017-03793 - 2013-02802 - 2014-04903
- 2017-01588 - 2013-02764 - 2012-00004
- 2017-01422 - 2012-04613 - 2012-06773
- 2016-04008 - 2015-01803 - 2012-01245
- 2015-04160 - 2008-01616 - 2016-00026
- 2015-02056 - 2013-03392 - 2012-03818
- 2015-01737 - 2012-05719
3
DESIGN BASIS DOCUMENTS
- DBD-1.09; Design Basis Documents for the Main Steam System; Revision 4
- DBD-2.07; Design Basis Document for Spent Fuel Pool Cooling System ; Revision 5
- DBD-5.01; Diesel and Auxiliary System; Revision 7
- DBD-5.03; Design Basis Document for Emergency Diesel Generator Performance Criteria;
Revision 9
- DBD-1.04; Design Basis Document for Chemical Volume Control System; Revision 7
DRAWINGS
- C-138; Containment Liner Support Trusses; Revision 9
- C-246; Reactor Building Safety Injection Tank Supports; Revision 4
- E-87, Sh.6; Schematic Diagram, CST Level and Alarm Indication; Revision 10
- E-100, Sh. 1; Schematic Diagram 480 V MCC Combination Starter and Feeders; Revision 28
- E-128, Sh.1; Schematic Diagram Charging Pump P55A Feeder Breaker Internals; Revision 5
- E-238, Sh. 1; Schematic Diagram Main Steam Isolation Valves; Revision 27
- E-257, Sh.1; Schematic Diagram Charging Pump P55A; Revision 24
- E-376, Sh.1; Conduit and Tray Plan for CST Instrument Line Freeze Protection; Revision 31
- E-679, Sh. 1; Schematic Diagram Diesel Oil Transfer Pumps; Revision 22
- E-897; Wiring Diagram, Freeze Protection Panel C100 & C100A; Revision 15
- M-202; P&ID Replacement Heat Tracing for CVC System; Revision 16
- M-205, Sh. 1; Connection Diagram SV-505A & B, Panel C-180 Piping & Instrument Diagram
Main Steam & Auxiliary Turbine Systems; Revision 94
- M-214, Sh. 1; Piping & Instrument Diagram, Lube Oil, Fuel Oil & Diesel Generator Systems;
Revision 81
- M-221 Sheet 2; Piping and Instrumentation Diagram Spent Fuel Pool Cooling System;
Revision 61
- M-214; Piping and Instrument Diagram Lube Oil, Fuel Oil and Diesel Generator Systems;
Revision 81
- M-221 SHT 2; Piping and Instrumentation Diagram Spent Fuel Pool Cooling System;
Revoision 61
- C-92; Auxiliary Building Fuel Pool Liner Plate Details; Revision 8
- C-111; Auxiliary Building Spent Fuel Pool Sections and Detail; Revision 6
- C-110; Auxiliary Building Spent Fuel Pool Sections and Detail; Revision 7
- M-207; 0002; Piping and Instrument Diagram, Auxiliary Feedwater System; Revision 41
- M-214; Piping and Instrument Diagram, Lube Oil, Fuel Oil, and Diesel Generator Systems;
Revision 41
- E-1 SHT A; Station Key Diagram; Revision 14
- E-4 SHT 1; Single Line Diagram 480 V Load Centers; Revision 45
- E-4 SHT 2; Single Line Diagram 480 V Load Centers; Revision 41
- E-5 SHT 1; Single Line Diagram 480 V Motor Control Centers; Revision 59
ENGINEERING CHANGES
- EC 58140; Install Permanent Shielding on Letdown Heat Exchanger E-58; Revision 0
- EC 58141; Install Shielding on Pressurizer Surge Line E-50A Platform; Revision 0
- EC 74734; Operability Input for Containment Dome Truss, Containment Spray and Safety
Injection Relative to CR-2017-5016
- EC 74877; Input for Operability of Containment Spray Supports HC44-R884, HC44-884.1,
HC44-R884.2, HC44-R884.3, HC44-R884.4, HC44-R884.5, and HC 44-R884.9; Revision 0
4
- EC 71766; Replacement of Diesel Fuel Oil Tank, T-10; Revision 0
- EA-SC-96-051-01; Fuel Oil Transfer Pump Replacement; Revision 1
- EC 56644; Diesel Generator Fuel Oil Tank to Comply with Michigan Fire Code; 05/18/2015
- EC 49797; Flex EC#22; Revision 0
- EC 47340; Flex EC#7; Revision 0
- EC 5000122470; Fast Bus Transfer Modification to Resolve TIA 2007-002; Revision 0
MISCELLANEOUS
- EDS Nuclear Report 02-0660-1087; Seismic Evaluation of Safety Injection Tank for Palisades
Nuclear Plant; Revision 1
- ENN-DC-152; Preparation, Revision, Review, and Approval of Design Basis Documents;
Revision 8
- Program SEP-AOV-PLP-001; Palisades Nuclear Power Plant Air Operated Valve Program
Entergy Nuclear Engineering Programs; Revision 1
- Specification No. C-173(Q); Technical Requirements for the Analysis and Design of
Safety-Related Pipe Supports; Revision 6
- M0120 0009; Laurence, RG Co. Inc. Information Bulletin for Safety Shut-Off 2-Way Manually
Reset Solenoid Valves; Revision 0
- VTD-0010-0140; General Electric Instructions for Polyphase Induction Motors; Revision 0
- VTD-0660-0052; Vendor Manual, Rosemount Pressure Transmitter; Revision 0
- PLP-RPT-12-00026; Maintenance Rule Scoping DocumentSpent Fuel Pool Cooling;
Revision 1
- PLP-RPT-12-00026; Maintenance Rule Scoping DocumentAuxiliary Building; Revision 1
- PLP-RPT-12-00026; Maintenance Rule Scoping DocumentChemical Volume Control
System; Revision 1
- PLP-RPT-12-00026; Maintenance Rule Scoping DocumentFuel Oil System ; Revision 1
- PLLP-ESPO-PBSO-VAS; Palisades Basic HVAC System Orientation; Revision 4
- Spent Fuel Pool System Health Report; 10/05/2017
- Chemical Volume ControlCharging/Letdown System Health Report; 10/20/2017
- System Health ReportSwitchyard System Q2-2017
- System Health Report2400 VAC System Q3-2017
- System Health Report480 VAC System
- L-HU-06-010; Response to GL 2006-002; 4/3/2006
- Response to RAI Regarding GL 2006-002; 1/29/2007
- SEP-MOV-PLP-001; MOV Program; Revision 2
- E0005-SH-0149-0000; Vendor ManualSiemens Vacuum Circuit Breakers
- Docket 50-265License DPR-20; Palisades Inservice Inspection Program Submittal of Relief
Request No. 14; Revision 1
- Palisades PlantAlternative to Defer Repair of Spent Fuel Pool Heat Exchanger E-53A
Nozzle Weld; 04/14/2000
- A-PAL-98-072; Lack of Procedural Guidance for Placing SFP Cooling on Emergency Power;
09/01/1998
- G-ME-A39; Spent Fuel Pool Cooling SystemSingle Failure Analysis; 11/10/1976
- EN-LI-119; Apparent Cause Evaluation for Historical Spent Fuel Pool Leakage Step Rise;
11/14/2010
- EAR-99-0081; CVCS Declassification; 3/29/1999
- Letter; Palisades PlantResolution of Unresolved Safety Issue (USI) A-46, Verification of
Seismic Qualification of Equipment in Operating Plants; 09/25/1998
5
OPERABILITY EVALUATIONS
- 2017-5016; Containment Dome Truss; 11/02/2017;
- 2017-5246; Containment Spray Pipe Support Straps; 11/14/2017
- EN-OP-104 Attachment 9.5; Condition Report Operability Evaluation 2017-02655 and
2017 -0266; 5/24/2017
- EN-OP-104 Attachment 9.5; Condition Report Operability Evaluation 2013-02855; 07/03/2013
PROCEDURES
- Procedure No. QO-37; Palisades Nuclear Plant Technical Specification Surveillance
Procedure Main Steam Isolation and Bypass Valve Testing; Revision 13
- ENN-DC-152; Preparation, Revision, Review, and Approval of Design Basis Documents;
Revision 8
- EM-04-58; Spent Fuel Pool METAMIC' Coupon Surveillance Program; Revision 2
- AOP-30; Loss of Shutdown Cooling; Revision 0
- AOP-35; Loss of Service Water; Revision 0
- AOP-41; Alternate Safe Shutdown Procedure; Revision 3
- SOP-24; Ventilation and Air Conditioning System; Revision 75
- EN-DC-150; Condition Monitoring of Maintenance Rule Structures; Revision 13
- RO-28; Control Room Envelope Positive Pressure; Revision 31
- RT-2-2; Control Room HVAC Heat Removal Capability; Revision 14
- MO-33; Control Room Ventilation Emergency Operation; Revision 26
- SOP-27; Fuel Pool System; Revision 72
- AOP-26; Loss of Spent Fuel Pool Cooling; Revision 3
- WI-SFP;O-01; Spent Fuel Pool Cooling Pump Oil Sample; Revision 1
- SOP-2B; Chemical and Volume Control Purification and Chemical Injection; Revision 53
- SOP-2A; Chemical Volume Control System Standard Design Process; Revision 87
- EN-DC-115; Engineering Change Process; Revision 21
- EN-DC-105; Configuration Management; Revision 4
- MO-8A-1; Emergency Diesel Generator 1-1; Revision 96
- SOP-22; Emergency Diesel Generators; Revision 74
- AOP-38; Acts of Nature; Revision 11
- CVCO-4; Periodic Test ProcedureCharging Pumps; Revision 11
- RT-71H; Spent Fuel System Class 3 Inservice Test; Revision 9
- WI-SFP-O-01; Spent Fuel Pool Cooling Pump Oil Sample; Revision 1
- DWO-1; Operator Daily/Weekly Items Modes 1,2,3, and 4; Revision 108
- MC-17; Fuel Oil Sampling; Revision 33
- SEP-PLP-IST-102; Inservice Testing of Selected Safety Related Pumps; Revision 3
- EN-DC-153; Preventative Maintenance Component Classification; Revision 15
- EN-DC-310; Predictive Maintenance Program; Revision 8
- SEP-VIB-PLP-001; Palisades Vibration Monitoring Program; Revision 2
- RO-112; Reactor Head/Pressurizer Vent Flow Check; Revision 11
- EN-LI-118; Causal Evaluation Process; Revision 24
- SOP-23, Att 9; Cold Weather Checklist-Electrical; Revision 58
- RI-125; CST Level Instrument Calibration; Revision 13
- EOP-3.0; Station Blackout Recovery; Revision 18
- ARP-7; Auxiliary Systems Scheme EK-11; Revision 101
- SPS-E-20; Maintenance for 2400V Siemens Switchgear; Revision 7
- SPS-E-28; Safeguards Transformer 1-1 Load Tap Transformer 1-1; Revision 8
- SPS-E-27; Inspection and Testing of Safeguards Transformer 1-1; Revision 8
6
- SOP-32; 345KV Switchyard Operating Procedure; Revision 38
PROGRAMS
- SEP-AOV-PLP-001; Palisades Nuclear Power Plant Air Operated Valve Program Entergy
Nuclear Engineering Programs; Revision 1
SURVEILLANCES/TESTING
- Fuel Oil Transfer Pump 18A IST Data; 2014-November 2017
- P-55A Vibration Data; 2014-November 2017
- RE-138; Calibration of Bus 1D Undervoltage and Time Delay Relays; Revision 15
WORK ORDERS/REQUEST
- WO 52655743; QO037 Main Steam Isolation and Bypass Valve Testing; 05/11/2017
- WO 52565880; QO037 Main Steam Isolation and Bypass Valve Testing; 10/10/2015
- WO 375934; QO037; Main Steam Isolation and Bypass Valve Testing; 06/21/2014
- WO 52751091; Emergency Diesel Fuel Oil Transfer Pump Test; 10/05/2017
- WO 00355289; P-18A, Troubleshoot and Correct Air In-Leakage; 08/23/2016
- WO 52574758; P-18A, Coupling Lubrication PM; 04/28/2016
- WO 52458340; RT-71H - Spent Fuel Pool System Class 3 Inservice Test; 09/28/2015
- WO 51628139; P-51B, Coupling Lubrication PM; 10/31/2017
- WO 5168138; P-51A, Coupling Lubrication PM; 10/31/2017
- WO 51627309; P-51B, Pump Bearing Oil Change; 10/31/2017
- WO 516273308; P-51A, Pump Bearing Oil Change; 10/31/2017
- WO 429141-01; RT-8C Engineered Safeguards EDG 1-1; 06/07/2017
- WO 327612-03; EDG 1-1 Voltage Regulator Replacement; 11/05/2007
- WO 52668400-01; Safeguards Transformer Load Tap Changer Maintenance; 09/10/2015
- WO 52578266; PM for Breaker 152-107; 02/04/2016;
- WO 52585445; DVR Time Surveillance 162-154; 03/21/2016
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LIST OF ACRONYMS USED
CAP Corrective Action Program
CFR Code of Federal Regulations
CR Condition Report
CST Condensate Storage Tank
EC Engineering Change
EDG Emergency Diesel Generator
IEEE Institute of Electrical and Electronic Engineers
IMC Inspection Manual Chapter
LERF Large Early Release Frequency
LOCA Loss-of-Coolant Accident
LOOP Loss of Off-Site Power
NCV Non-Cited Violation
NPSH Net Positive Suction Head
NRC U.S. Nuclear Regulatory Commission
RG Regulatory Guide
SIT Safety Injection Tank
UFSAR Updated Final Safety Analysis Report
URI Unresolved Item
8