ML17131A207
| ML17131A207 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 05/11/2017 |
| From: | Jamnes Cameron Reactor Projects Region 3 Branch 4 |
| To: | Hamilton D FirstEnergy Nuclear Operating Co |
| References | |
| IR 2017001 | |
| Download: ML17131A207 (54) | |
See also: IR 05000440/2017001
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION III
2443 WARRENVILLE RD. SUITE 210
LISLE, IL 60532-4352
May 11, 2017
Mr. David B. Hamilton
Site Vice President
FirstEnergy Nuclear Operating Company
Perry Nuclear Power Plant
Mail Stop A-PY-A290
P.O. Box 97, 10 Center Road
Perry, OH 44081-0097
SUBJECT: PERRY NUCLEAR POWER PLANTNRC INTEGRATED INSPECTION REPORT
Dear Mr. Hamilton:
On March 31, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed a baseline
inspection at your Perry Nuclear Power Plant. On April 13, 2016, the NRC inspectors discussed
the results of this inspection with Mr. F. Payne and other members of your staff. The enclosed
report represents the results of this inspection.
Based on the results of this inspection, the NRC has identified one issue that was evaluated
under the risk significance determination process as having very low safety significance
(Green). The NRC has also determined that a violation is associated with this issue. Because
the licensee initiated condition reports to address this issue, this violation is being treated as a
Non-Cited Violation (NCV), consistent with Section 2.3.2 of the Enforcement Policy. The NCV is
described in the subject inspection report. Furthermore, the inspectors documented a licensee-
identified violation which was determined to be of very low safety significance in this report. The
NRC is treating this violation as an NCV consistent with Section 2.3.2.a of the Enforcement
Policy.
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 III; the Director, Office of Enforcement; and the
NRC resident inspector at the Perry Nuclear Power Plant.
D. Hamilton -2-
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/
Jamnes Cameron, Chief
Branch 4
Division of Reactor Projects
Docket No. 50-440
License No. NPF-58
Enclosure:
Inspection Report 05000440/2017001
cc: Distribution via LISTSERV
D. Hamilton -3-
Letter to David Hamilton from Jamnes Cameron dated May 11, 2017
SUBJECT: PERRY NUCLEAR POWER PLANTNRC INTEGRATED INSPECTION REPORT
DISTRIBUTION:
RidsNrrPMPerry Resource
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RidsNrrDirsIrib Resource
Cynthia Pederson
DRPIII
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ROPreports.Resource@nrc.gov
ADAMS Accession Number: ML17131A207
OFFICE RIII
NAME JCameron:bw
DATE 05/ /2017
OFFICIAL RECORD COPY
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No: 50-440
License No: NPF-58
Report No: 05000440/2017001
Licensee: FirstEnergy Nuclear Operating Company (FENOC)
Facility: Perry Nuclear Power Plant
Location: North Perry, Ohio
Dates: January 1 through March 31, 2017
Inspectors: R. Elliott, Acting Senior Resident Inspector
J. Nance, Acting Senior Resident Inspector
M. Doyle, Acting Resident Inspector
V. Meghani, Reactor Inspector
S. Bell, Health Physicist
V. Meyers, Senior Health Physicist
N. Féliz Adorno, Senior Reactor Inspector
M. Jones, Reactor Inspector
Approved by: J. Cameron, Chief
Branch 4
Division of Reactor Projects
Enclosure
TABLE OF CONTENTS
SUMMARY .................................................................................................................................... 2
REPORT DETAILS ....................................................................................................................... 3
Summary of Plant Status ........................................................................................................... 3
1. REACTOR SAFETY ............................................................................................ 3
1R04 Equipment Alignment (71111.04) ........................................................................ 3
1R05 Fire Protection (71111.05) ................................................................................... 4
1R08 Inservice Inspection Activities (71111.08G) ......................................................... 6
1R11 Licensed Operator Requalification Program (71111.11) ...................................... 8
1R12 Maintenance Effectiveness (71111.12) ................................................................ 9
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13) ........ 10
1R15 Operability Determinations and Functional Assessments (71111.15) ............... 10
1R18 Plant Modifications (71111.18) .......................................................................... 11
1R19 Post-Maintenance Testing (71111.19) ............................................................... 12
1R20 Outage Activities (71111.20) .............................................................................. 13
1R22 Surveillance Testing (71111.22) ........................................................................ 14
1EP6 Drill Evaluation (71114.06)................................................................................. 15
2. RADIATION SAFETY ........................................................................................ 16
2RS1 Radiological Hazard Assessment and Exposure Controls (71124.01) .............. 16
2RS2 Occupational As-Low-As-Reasonably-Achievable Planning and Controls
(71124.02).......................................................................................................... 20
4OA1 Performance Indicator Verification (71151) ....................................................... 22
4OA2 Identification and Resolution of Problems (71152) ............................................ 23
4OA5 Other Activities ................................................................................................... 27
4OA6 Management Meetings....................................................................................... 31
4OA7 Licensee-Identified Violations ............................................................................ 31
SUPPLEMENTAL INFORMATION............................................................................................ 1
Key Points of Contact ................................................................................................................ 1
List of Items Opened, Closed, and Discussed........................................................................... 2
List of Documents Reviewed ..................................................................................................... 3
List of Acronyms Used ............................................................................................................ 13
SUMMARY
Inspection Report (IR) 05000440/2017001, 01/01/2017 - 03/31/2017, Perry Nuclear Power
Plant; Routine Integrated Inspection Report.
This report covers a 3-month period of inspection by resident inspectors and announced
baseline inspections by regional inspectors. The significance of inspection findings is indicated
by their color (i.e., greater than Green, or Green, White, Yellow, Red) and determined using
Inspection Manual Chapter (IMC) 0609, "Significance Determination Process," dated
April 29, 2015. 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.
Cornerstone: Mitigating Systems
Green. A finding of very-low safety significance and associated NCV of TS 5.4,
Procedures, was identified by the inspectors for the failure to implement procedures for
combating a loss of shutdown cooling (SDC). Specifically, the licensee failed to
implement its procedure for combating a loss of SDC resulting from emergency service
water (ESW) inoperability and during high decay heat load. This finding was entered
into the licensees Corrective Action Program to perform analyses for various conditions
to identify available alternate methods of decay heat removal and provide associated
procedural guidance.
The performance deficiency was determined to be more-than-minor because it was
associated with the Mitigating Systems cornerstone attribute of design control and
affected the cornerstone objective of ensuring the availability, reliability, and capability of
mitigating systems to respond to initiating events to prevent undesirable consequences.
The finding screened as very-low safety significance (Green) because it was a design
deficiency that did not impact the operability or Probabilistic Risk Assessment
functionality of any mitigating structures, systems, and components. The inspectors did
not identify a cross-cutting aspect associated with this finding because it did not reflect
current performance due to the age of the performance deficiency.
(Section 4OA5.1.b(1))
Licensee-Identified Violations
A Violation of very low safety significance was identified by the licensee and has been
reviewed by the NRC. Corrective actions taken or planned by the licensee have been
entered into the licensees corrective action program (CAP). This violation and CAP
tracking number is listed in Section 4OA7 of this report.
2
REPORT DETAILS
Summary of Plant Status
The plant began the inspection period at 98 percent power, due to end-of-core life prior to
refueling outage (RFO) 1R16. The operators performed minor power reductions during this
inspection period to support routine surveillances while the plant continued to coastdown until
March 5, when at 12:01 a.m., the plant disconnected from the grid and was shut down for RFO
1R16. The plant remained in RFO 1R16 through the end of the quarter.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and
1R04 Equipment Alignment (71111.04)
.1 Quarterly Partial System Walkdowns
a. Inspection Scope
The inspectors performed partial system walkdowns of the following risk-significant
systems:
- standby liquid control system (SLC) B;
- division 1 emergency diesel generator (EDG);
- control room emergency ventilation system B; and
- Unit 2 startup transformer.
The inspectors selected these systems based on their risk significance relative to the
Reactor Safety Cornerstones at the time they were inspected. The inspectors attempted
to identify any discrepancies that could impact the function of the system and, therefore,
potentially increase risk. The inspectors reviewed applicable operating procedures,
system diagrams, Updated Safety Analysis Report (USAR), Technical Specification (TS)
requirements, outstanding work orders (WOs), condition reports, and the impact of
ongoing work activities on redundant trains of equipment in order to identify conditions
that could have rendered the systems incapable of performing their intended functions.
The inspectors also walked down accessible portions of the systems to verify system
components and support equipment were aligned correctly and operable.
The inspectors examined the material condition of the components and observed
operating parameters of equipment to verify that there were no obvious deficiencies.
The inspectors also verified that the licensee had properly identified and resolved
equipment alignment problems that could cause initiating events or impact the capability
of mitigating systems or barriers and entered them into the CAP with the appropriate
significance characterization. Documents reviewed are listed in the Attachment to this
report.
These activities constituted four partial system walkdown samples as defined in
inspection procedure (IP) 71111.04-05.
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b. Findings
No findings were identified.
.2 Semi-Annual Complete System Walkdown
a. Inspection Scope
On March 6, 2017, the inspectors completed a system alignment inspection of the high
pressure core spray system to verify the functional capability of the system. This system
was selected because it was considered both safety significant and risk significant in the
licensees probabilistic risk assessment. The inspectors walked down the system to
review mechanical and electrical equipment lineups; electrical power availability; system
pressure and temperature indications, as appropriate; component labeling; component
lubrication; component and equipment cooling; hangers and supports; operability of
support systems; and to ensure that ancillary equipment or debris did not interfere with
equipment operation. A review of a sample of past and outstanding WOs was
performed to determine whether any deficiencies significantly affected the system
function. In addition, the inspectors reviewed the CAP database to ensure that system
equipment alignment problems were being identified and appropriately resolved.
Documents reviewed are listed in the Attachment to this report.
These activities constituted one complete system walkdown sample as defined in
IP 71111.04-05.
b. Findings
No findings were identified.
1R05 Fire Protection (71111.05)
.1 Routine Resident Inspector Tours (71111.05Q)
a. Inspection Scope
The inspectors conducted fire protection walkdowns which were focused on availability,
accessibility, and the condition of firefighting equipment in the following risk-significant
plant areas:
- fire zone OCC-1A; control complex 57410;
- fire zone 1RB-1C; containment building, 599, 620, 642, 652, 6647, 599, and
689 elevations;
- fire zone OFH-3; fuel handling building; 6206; and
- fire zone 1RB-1C-1B; drywell; 5836, 599, 6206, and 6368.
The inspectors reviewed areas to assess whether the licensee had implemented a fire
protection program that adequately controlled combustibles and ignition sources within
the plant, effectively maintained fire detection and suppression capability, maintained
passive fire protection features in good material condition, and implemented adequate
compensatory measures for out-of-service, degraded or inoperable fire protection
equipment, systems, or features in accordance with the licensees fire plan.
4
The inspectors selected fire areas based on their overall contribution to internal fire risk
as documented in the plants Individual Plant Examination of External Events with later
additional insights, their potential to impact equipment which could initiate or mitigate a
plant transient, or their impact on the plants ability to respond to a security event.
Using the documents listed in the Attachment to this report, the inspectors verified that
fire hoses and extinguishers were in their designated locations and available for
immediate use; that fire detectors and sprinklers were unobstructed; that transient
material loading was within the analyzed limits; and fire doors, dampers, and penetration
seals appeared to be in satisfactory condition. The inspectors also verified that minor
issues identified during the inspection were entered into the licensees CAP.
Documents reviewed are listed in the Attachment to this report.
These activities constituted five quarterly fire protection inspection samples as defined in
IP 71111.05-05.
b. Findings
No findings were identified.
.2 Annual Fire Protection Drill Observation (71111.05A)
a. Inspection Scope
On February 2, 2017 and February 6, 2017, the inspectors observed two fire brigade
activation unannounced drills. On February 27, 2017, the inspectors observed the
licensee response to a fire on Screen Wash Pump A in the Service Water Building.
Based on these observations, the inspectors evaluated the readiness of the plant fire
brigade to fight fires. The inspectors verified that the licensee staff identified
deficiencies, openly discussed them in a self-critical manner at the drill debrief, and took
appropriate corrective actions. Specific attributes evaluated were:
- proper wearing of turnout gear and self-contained breathing apparatus;
- proper use and layout of fire hoses;
- employment of appropriate firefighting techniques;
- sufficient firefighting equipment brought to the scene;
- effectiveness of fire brigade leader communications, command, and control;
- search for victims and propagation of the fire into other plant areas;
- smoke removal operations;
- utilization of pre-planned strategies;
- adherence to the pre-planned drill scenario; and
- drill objectives.
Documents reviewed are listed in the Attachment to this report.
These activities constituted one annual fire protection inspection sample as defined in
IP 71111.05-05.
b. Findings
No findings were identified.
5
1R08 Inservice Inspection Activities (71111.08G)
From March 6, 2017 through March 10, 2017, the inspectors conducted a review of
the implementation of the licensees Inservice Inspection (ISI) Program for monitoring
degradation of the reactor coolant system, risk-significant piping and components,
and containment systems.
The ISIs described in Sections 1R08.1 and 1R08.5 below constituted one inspection
sample as defined in Inspection Procedure 71111.08-05.
.1 Piping Systems Inservice Inspection
a. Inspection Scope
The inspectors either observed or reviewed the following non-destructive examinations
mandated by the American Society for Mechanical Engineers (ASME),Section XI Code,
to evaluate compliance with the ASME Code Section XI and Section V requirements,
and whether any indications and defects were detected to determine whether these were
dispositioned in accordance with the ASME Code or an NRC approved alternative
requirement.
- ultrasonic examination (UT) of residual heat removal (RHR), 12 inch
valve-to-pipe weld, 1E12-F053A;
- visual-3 examination (VT-3) of mechanical snubber support, 1E21-H0004;
- UT of reactor pressure vessel (RPV) head studs, report UT-17-E005;
- magnetic Particle examination of piping support welded attachment,
- VT-3, examination of low pressure core spray (LPCS) system variable spring
support, 1E21-H0025; and
- VT-3, examination of RHR system rigid strut support, IE12 H0633.
The inspectors reviewed the following examinations completed during the previous
outage with relevant/recordable conditions/indications accepted for continued service to
determine whether acceptance was in accordance with the ASME Code Section XI, or
an NRC-approved alternative:
- disposition for indication detected during automated UT of shroud support plate
to reactor vessel wall weld H-9 (WO 200567907);
- indication (VT-3) disposition rejected during variable spring hanger examination
of 1E22-H0071 (condition report CR 2015-02652); and
- indication (VT-3) disposition rejected during examination of mechanical snubber
support 1E12-H0765 (WO 200569179, condition report CR 2015-03374).
6
The inspectors reviewed records for the following pressure boundary weld repairs
completed for a risk-significant system during the last outage to determine whether the
licensee applied the pre-service non-destructive examinations, and acceptance criteria
required by the Construction Code, and/or the NRC-approved Code relief request.
Additionally, the inspectors reviewed the welding procedure specification and supporting
weld procedure qualification records to determine whether the weld procedure was
qualified in accordance with the requirements of the Construction Code and the ASME
Code,Section IX.
- replace 1B33F0029 drain valve, reactor recirculation system (WO 200391180);
and
- re-Installation of existing test port plug after fiberscopic inspection of valve
internals PY-IN27F0559B (WO 200565864).
b. Findings
No findings were identified.
.2 Reactor Pressure Vessel Upper Head Penetration Inspection Activities (Not Applicable)
.3 Boric Acid Corrosion Control (Not Applicable)
.4 Steam Generator Tube Inspection Activities (Not Applicable)
.5 Identification and Resolution of Problems
a. Inspection Scope
The inspectors performed a review of ISI-related problems entered into the licensees
CAP, and conducted interviews with licensee staff to determine whether the licensee
had:
- established an appropriate threshold for identifying ISI-related problems;
- performed a root cause (if applicable) and taken appropriate corrective actions;
and
- evaluated operating experience and industry generic issues related to ISI and
pressure boundary integrity.
The inspectors performed these reviews to evaluate compliance with Title 10 of the
Code of Federal Regulations, (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective
Action, requirements. The corrective action documents reviewed by the inspectors are
listed in the Attachment to this report.
b. Findings
No findings were identified.
7
1R11 Licensed Operator Requalification Program (71111.11)
.1 Resident Inspector Quarterly Review of Licensed Operator Requalification (71111.11Q)
a. Inspection Scope
On January 23, 2017, the inspectors observed a crew of licensed operators in the plants
simulator during licensed operator requalification training. The inspectors verified that
operator performance was adequate, evaluators were identifying and documenting crew
performance problems, and that training was being conducted in accordance with
licensee procedures. The inspectors evaluated the following areas:
- licensed operator performance;
- crews clarity and formality of communications;
- ability to take timely actions in the conservative direction;
- prioritization, interpretation, and verification of annunciator alarms;
- correct use and implementation of abnormal and emergency procedures;
- control board manipulations;
- oversight and direction from supervisors; and
- ability to identify and implement appropriate TS actions and Emergency Plan
(EP) actions and notifications.
The crews performance in these areas was compared to pre-established operator action
expectations and successful critical task completion requirements. Documents reviewed
are listed in the Attachment to this report.
This inspection constituted one quarterly licensed operator requalification program
simulator sample as defined in IP 71111.11-05.
b. Findings
No findings were identified.
.2 Resident Inspector Quarterly Observation During Periods of Heightened Activity or Risk
(71111.11Q)
a. Inspection Scope
On March 4 and 5, 2017, the inspectors observed the shutdown of Perry and entry into
RFO16. This was an activity that required heightened awareness or was related to
increased risk. The inspectors evaluated the following areas:
- licensed operator performance;
- crews clarity and formality of communications;
- ability to take timely actions in the conservative direction;
- prioritization, interpretation, and verification of annunciator alarms (if applicable);
- correct use and implementation of procedures;
- control board (or equipment) manipulations;
- oversight and direction from supervisors; and
- ability to identify and implement appropriate TS actions and EP actions and
notifications (if applicable).
8
The performance in these areas was compared to pre-established operator action
expectations, procedural compliance and task completion requirements. Documents
reviewed are listed in the Attachment to this report.
This inspection constituted one quarterly licensed operator heightened activity/risk
sample as defined in IP 71111.11-05.
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness (71111.12)
.1 Routine Quarterly Evaluations
a. Inspection Scope
The inspectors evaluated degraded performance issues involving the following
risk-significant systems:
- metal clad 5 kilovolts switchgear; and
- neutron monitoring.
The inspectors reviewed events such as where ineffective equipment maintenance had
or could have resulted in valid or invalid automatic actuations of engineered safeguard
systems and independently verified the licensee's actions to address system
performance or condition problems in terms of the following:
- implementing appropriate work practices;
- identifying and addressing common cause failures;
- scoping of systems in accordance with 10 CFR 50.65(b) of the maintenance rule;
- characterizing system reliability issues for performance;
- charging unavailability for performance;
- trending key parameters for condition monitoring;
- ensuring 10 CFR 50.65(a)(1) or (a)(2) classification or re-classification; and
- verifying appropriate performance criteria for structures, systems, and
components (SSCs)/functions classified as (a)(2), or appropriate and adequate
goals and corrective actions for systems classified as (a)(1).
The inspectors assessed performance issues with respect to the reliability, availability,
and condition monitoring of the system. In addition, the inspectors verified maintenance
effectiveness issues were entered into the CAP with the appropriate significance
characterization. Documents reviewed are listed in the Attachment to this report.
This inspection constituted two quarterly maintenance effectiveness samples as defined
in IP 71111.12-05.
b. Findings
No findings were identified.
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1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)
.1 Maintenance Risk Assessments and Emergent Work Control
a. Inspection Scope
The inspectors reviewed the licensee's evaluation and management of plant risk for the
maintenance and emergent work activities affecting risk-significant and safety-related
equipment listed below to verify that the appropriate risk assessments were performed
prior to removing equipment for work:
- high pressure core spray (HPCS) condensate storage tank test pressure
instrument root valve weld repair;
- Unit 2 startup transformer out of service during high winds;
- shutdown risk yellow with Division 1 EDG inoperable;
- shutdown risk yellow during replacement of B ESW pump; and
- Unit 1 startup transformer out of service with switchyard breakers S610 and S620
out of service.
These activities were selected based on their potential risk significance relative to the
Reactor Safety Cornerstones. As applicable for each activity, the inspectors verified that
risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate
and complete. When emergent work was performed, the inspectors verified that the
plant risk was promptly reassessed and managed. The inspectors reviewed the scope
of maintenance work, discussed the results of the assessment with the licensee's
probabilistic risk analyst or shift technical advisor, and verified plant conditions were
consistent with the risk assessment. The inspectors also reviewed TS requirements and
walked down portions of redundant safety systems, when applicable, to verify risk
analysis assumptions were valid and applicable requirements were met.
Documents reviewed during this inspection are listed in the Attachment to this report.
These maintenance risk assessments and emergent work control activities constituted
five samples as defined in IP 71111.13-05.
b. Findings
No findings were identified.
1R15 Operability Determinations and Functional Assessments (71111.15)
.1 Operability Evaluations
a. Inspection Scope
The inspectors reviewed the following issues:
- standby liquid control (SLC) B squib valve operability;
- reactor core isolation cooling (RCIC) operability with room watertight door found
open and unattended;
- underdrain and gravity discharge system rock salt intrusion functionality
assessment;
10
- operations with potential of draining the reactor vessel (OPDRV) requirements for
control room emergency recirculation and control room heating, ventilation, and
air conditioning inoperabilities; and
- residual heat removal (RHR) A and B operability due to a degraded snubber.
The inspectors selected these potential operability issues based on the risk significance
of the associated components and systems. The inspectors evaluated the technical
adequacy of the evaluations to ensure that TS operability was properly justified and the
subject component or system remained available such that no unrecognized increase in
risk occurred. The inspectors compared the operability and design criteria in the
appropriate sections of the TS and Updated Safety Analysis Report (USAR) to the
licensees evaluations to determine whether the components or systems were operable.
Where compensatory measures were required to maintain operability, the inspectors
determined whether the measures in place would function as intended and were
properly controlled. The inspectors determined, where appropriate, compliance with
bounding limitations associated with the evaluations. Additionally, the inspectors
reviewed a sampling of corrective action documents to verify that the licensee was
identifying and correcting any deficiencies associated with operability evaluations.
Documents reviewed are listed in the Attachment to this report.
This operability inspection constituted five sample as defined in IP 71111.15-05.
b. Findings
No findings were identified.
1R18 Plant Modifications (71111.18)
.1 Plant Modifications
a. Inspection Scope
The inspectors reviewed the following modification(s):
- ECP 16-0178-000; Diesel Generator Ventilation Bypass Switch Modification.
The inspectors reviewed the configuration changes and associated 10 CFR 50.59 safety
evaluation screening against the design basis, the USAR, and the TS, as applicable, to
verify that the modification did not affect the operability or availability of the affected
system(s). The inspectors, as applicable, observed ongoing and completed work
activities to ensure that the modifications were installed as directed and consistent with
the design control documents; the modifications operated as expected; post-modification
testing adequately demonstrated continued system operability, availability, and reliability;
and that operation of the modifications did not impact the operability of any interfacing
systems. As applicable, the inspectors verified that relevant procedure, design, and
licensing documents were properly updated. Lastly, the inspectors discussed the plant
modification with operations, engineering, and training personnel to ensure that the
individuals were aware of how the operation with the plant modification in place could
impact overall plant performance. Documents reviewed are listed in the Attachment to
this report.
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This inspection constituted one permanent plant modification sample as defined in
IP 71111.18-05.
b. Findings
No findings were identified.
1R19 Post-Maintenance Testing (71111.19)
.1 Post-Maintenance Testing
a. Inspection Scope
The inspectors reviewed the following post-maintenance (PM) activities to verify that
procedures and test activities were adequate to ensure system operability and functional
capability:
- Unit 2 start-up transformer replacement PM test.
These activities were selected based upon the structure, system, or component's ability
to impact risk. The inspectors evaluated these activities for the following (as applicable):
the effect of testing on the plant had been adequately addressed; testing was adequate
for the maintenance performed; acceptance criteria were clear and demonstrated
operational readiness; test instrumentation was appropriate; tests were performed as
written in accordance with properly reviewed and approved procedures; equipment was
returned to its operational status following testing (temporary modifications or jumpers
required for test performance were properly removed after test completion); and test
documentation was properly evaluated. The inspectors evaluated the activities against
TSs, the UFSAR, Title 10 of the Code of Federal Regulations (10 CFR) Part 50
requirements, licensee procedures, and various NRC generic communications to ensure
that the test results adequately ensured that the equipment met the licensing basis and
design requirements. In addition, the inspectors reviewed corrective action documents
associated with post-maintenance tests to determine whether the licensee was
identifying problems and entering them in the CAP and that the problems were being
corrected commensurate with their importance to safety. Documents reviewed are listed
in the Attachment to this report.
This inspection constituted six post-maintenance testing sample as defined in
IP 71111.19-05.
b. Findings
No findings were identified.
12
1R20 Outage Activities (71111.20)
.1 Refueling Outage Activities
a. Inspection Scope
The inspectors reviewed the Outage Safety Plan (OSP) and contingency plans for the
RFO 1R16 to confirm that the licensee had appropriately considered risk, industry
experience, and previous site-specific problems in developing and implementing a plan
that assured maintenance of defense-in-depth. During the RFO, the inspectors
observed portions of the shutdown and cooldown processes and monitored licensee
controls over the outage activities listed below:
- licensee configuration management, including maintenance of defense-in-depth
commensurate with the OSP for key safety functions and compliance with the
applicable TS when taking equipment out of service;
- implementation of clearance activities and confirmation that tags were properly
hung and equipment appropriately configured to safely support the work or
testing;
- installation and configuration of reactor coolant pressure, level, and temperature
instruments to provide accurate indication, accounting for instrument error;
- controls over the status and configuration of electrical systems to ensure that
TS and OSP requirements were met, and controls over switchyard activities;
- monitoring of decay heat removal processes, systems, and components;
- controls to ensure that outage work was not impacting the ability of the operators
to operate the spent fuel pool cooling system;
- reactor water inventory controls including flow paths, configurations, and
alternative means for inventory addition, and controls to prevent inventory loss;
- controls over activities that could affect reactivity;
- maintenance of secondary containment as required by TS;
- licensee fatigue management, as required by 10 CFR 26, Subpart I;
- refueling activities, including fuel handling and sipping to detect fuel assembly
leakage;
- startup and ascension to full power operation, tracking of startup prerequisites,
walkdown of the drywell (primary containment) to verify that debris had not been
left which could block emergency core cooling system suction strainers, and
reactor physics testing; and
- licensee identification and resolution of problems related to RFO activities.
Documents reviewed are listed in the Attachment to this report.
This inspection constituted one RFO sample as defined in IP 71111.20-05.
b. Findings
No findings were identified.
13
1R22 Surveillance Testing (71111.22)
.1 Surveillance Testing
a. Inspection Scope
The inspectors reviewed the test results for the following activities to determine whether
risk-significant systems and equipment were capable of performing their intended safety
function and to verify testing was conducted in accordance with applicable procedural
and TS requirements:
- SVI-C51-T0050-G; Osculating Power Range Monitor (OPRM) Channel G
Calibration for 1C51-K603G (routine);
- SVI-C41-T2001-B; SLC B Pump and Valve Operability Test (routine);
1B33A-CB4A and 1B33A-CB4B; dated March 4, 2017 (routine);
(ISO valve);
- SVI-D23-T2002A; Containment Atmosphere Monitoring Train A Isolation Valves
Seat Leakage and Position Indication Test; Revision 4 (ISO valve); and
- FTI-F0031; Volumetrics and FENOC Leak Rate Monitors Testing Instruction;
Revision 4 (routine).
The inspectors observed in-plant activities and reviewed procedures and associated
records to determine the following:
- did preconditioning occur;
- the effects of the testing were adequately addressed by control room personnel
or engineers prior to the commencement of the testing;
- acceptance criteria were clearly stated, demonstrated operational readiness, and
were consistent with the system design basis;
- plant equipment calibration was correct, accurate, and properly documented;
- as-left setpoints were within required ranges; and the calibration frequency was
in accordance with TSs, the UFSAR, procedures, and applicable commitments;
- measuring and test equipment calibration was current;
- test equipment was used within the required range and accuracy; applicable
prerequisites described in the test procedures were satisfied;
- test frequencies met TS requirements to demonstrate operability and reliability;
tests were performed in accordance with the test procedures and other
applicable procedures; jumpers and lifted leads were controlled and restored
where used;
- test data and results were accurate, complete, within limits, and valid;
- test equipment was removed after testing;
- where applicable for in-service testing activities, testing was performed in
accordance with the applicable version of Section XI, American Society of
Mechanical Engineers code, and reference values were consistent with the
system design basis;
14
- where applicable, test results not meeting acceptance criteria were addressed
with an adequate operability evaluation or the system or component was
declared inoperable;
- where applicable for safety-related instrument control surveillance tests,
reference setting data were accurately incorporated in the test procedure;
- where applicable, actual conditions encountering high resistance electrical
contacts were such that the intended safety function could still be accomplished;
- prior procedure changes had not provided an opportunity to identify problems
encountered during the performance of the surveillance or calibration test;
- equipment was returned to a position or status required to support the
performance of its safety functions; and
- all problems identified during the testing were appropriately documented and
dispositioned in the CAP.
Documents reviewed are listed in the Attachment to this report.
This inspection constituted four routine surveillance testing samples, one in-service test
sample, and two containment isolation valve samples as defined in IP 71111.22,
Sections-02 and-05.
b. Findings
No findings were identified.
1EP6 Drill Evaluation (71114.06)
.1 Training Observation
a. Inspection Scope
The inspectors observed a simulator training evolution for licensed operators on
January 23, 2017, which required emergency plan implementation by a licensee
operations crew. This evolution was planned to be evaluated and included performance
indicator data regarding drill and exercise performance. The inspectors observed event
classification and notification activities performed by the crew. The inspectors also
attended the post-evolution critique for the scenario. The focus of the inspectors
activities was to note any weaknesses and deficiencies in the crews performance and
ensure that the licensee evaluators noted the same issues and entered them into the
corrective action program. As part of the inspection, the inspectors reviewed the
scenario package and other documents listed in the Attachment to this report.
This inspection of the licensees training evolution with emergency preparedness drill
aspects constituted one sample as defined in IP 71114.06-06.
b. Findings
No findings were identified.
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2. RADIATION SAFETY
Cornerstones: Public Radiation Safety and Occupational Radiation Safety
2RS1 Radiological Hazard Assessment and Exposure Controls (71124.01)
.1 Radiological Hazard Assessment (02.02)
a. Inspection Scope
The inspectors assessed the licensees current and historic isotopic mix, including alpha
emitters and other hard-to-detect radionuclides. The inspectors evaluated whether
survey protocols were reasonable to identify the magnitude and extent of the radiological
hazards.
The inspectors determined whether there have been changes to plant operations since
the last inspection that may have resulted in a significant new radiological hazard for
onsite individuals. The inspectors evaluated whether the licensee assessed the
potential impact of these changes and implemented periodic monitoring, as appropriate,
to detect and quantify the radiological hazard. The inspectors reviewed the last two
radiological surveys from selected plant areas and evaluated whether the thoroughness
and frequency of the surveys were appropriate for the given radiological hazard.
The inspectors conducted walkdowns of the facility, including radioactive waste
processing, storage, and handling areas to evaluate material conditions and performed
independent radiation measurements as needed to verify conditions were consistent
with documented radiation surveys.
The inspectors assessed the adequacy of pre-work surveys for select radiologically
risk-significant work activities.
The inspectors evaluated the radiological survey program to determine whether hazards
were properly identified. The inspectors discussed procedures, equipment, and
performance of surveys with radiation protection staff and assessed whether technicians
were knowledgeable about when and how to survey areas for various types of
radiological hazards.
The inspectors reviewed work in potential airborne areas to assess whether air samples
were being taken appropriately for their intended purpose and reviewed various survey
records to assess whether the samples were collected and analyzed appropriately. The
inspectors also reviewed the licensees program for monitoring contamination which has
the potential to become airborne.
These inspection activities constituted one complete sample as defined in
IP 71124.01-05.
b. Findings
No findings were identified.
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.2 Instructions to Workers (02.03)
a. Inspection Scope
The inspectors reviewed select radiation work permits (RWPs) used to access high
radiation areas and evaluated the specified work control instructions or control barriers.
The inspectors also assessed whether workers where made aware of the work
instructions and area dose rates.
The inspectors reviewed electronic alarming dosimeter dose and dose rate alarm
setpoint methodology. For selected electronic alarming dosimeter occurrences, the
inspectors assessed the workers response to the alarm, the licensees evaluation of the
alarm, and any follow-up investigations.
The inspectors reviewed the licensees methods for informing workers of changes in
plant operations or radiological conditions that could significantly impact their
occupational dose.
The inspectors reviewed the labeling of select containers of licensed radioactive material
that could cause unplanned or inadvertent exposure to workers.
These inspection activities constituted one complete sample as defined in
IP 71124.01-05.
b. Findings
No findings were identified.
.3 Contamination and Radioactive Material Control (02.04)
a. Inspection Scope
The inspectors observed locations where the licensee monitors material leaving the
radiologically controlled area (RCA) and assessed the methods used for control, survey,
and release of material from these areas. As available, the inspectors observed health
physics personnel surveying and releasing material for unrestricted use.
The inspectors observed workers leaving the RCA and assessed their use of tool and
personal contamination monitors and reviewed the licensees criteria for use of the
monitors.
The inspectors assessed whether instrumentation was used at its typical sensitivity
levels based on appropriate counting parameters or whether the licensee had
established a de facto release limit.
The inspectors selected several sealed sources from the licensees inventory records
and assessed whether the sources were accounted for and verified to be intact. The
inspectors also evaluated whether any transactions, since the last inspection, involving
nationally tracked sources were reported in accordance with Title 10 CFR, Part 20.2207.
These inspection activities constituted one complete sample as defined in
IP 71124.01-05.
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b. Findings
No findings were identified.
.4 Radiological Hazards Control and Work Coverage (02.05)
a. Inspection Scope
The inspectors evaluated ambient radiological conditions during tours of the facility. The
inspectors assessed whether the conditions were consistent with applicable posted
surveys, radiation work permits (RWPs), and worker briefings.
The inspectors evaluated the adequacy of radiological controls, such as required
surveys, radiation protection job coverage, and contamination controls. The inspectors
evaluated the licensees use of electronic alarming dosimeters in high noise areas as
high radiation area monitoring devices.
The inspectors assessed whether radiation monitoring devices were placed on the
individuals body consistent with licensee procedures. The inspectors assessed whether
the dosimeter was placed in the location of highest expected dose or that the licensee
properly employed a NRC-approved method of determining effective dose equivalent.
The inspectors reviewed the application of dosimetry to effectively monitor exposure to
personnel in work areas with significant dose rate gradients.
For select airborne area RWPs, the inspectors reviewed airborne radioactivity controls
and monitoring, the potential for significant airborne levels, containment barrier integrity,
and temporary filtered ventilation system operation.
The inspectors examined the licensees physical and programmatic controls for highly
activated or contaminated materials stored within pools and assessed whether
appropriate controls were in place to preclude inadvertent removal of these materials
from the pool.
These inspection activities constituted one complete sample as defined in
IP 71124.01-05.
b. Findings
No findings were identified.
.5 High Radiation Area and Very High Radiation Area Controls (02.06)
a. Inspection Scope
The inspectors observed posting and physical controls for high radiation areas (HRAs)
and very HRAs to assess adequacy.
The inspectors conducted a selective inspection of posting and physical controls for
HRAs and very HRAs to assess conformance with performance indicators.
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The inspectors reviewed procedural changes to assess the adequacy of access controls
for high and very HRAs to determine whether procedural changes substantially reduced
the effectiveness and level of worker protection.
The inspectors assessed the controls for HRAs with greater than 1 rem/hour and areas
with the potential to become HRAs greater than 1 rem/hour for compliance with TS and
procedures.
The inspectors assessed the controls for very HRAs and areas with the potential to
become very HRAs. The inspectors also assessed whether individuals were unable to
gain unauthorized access to these areas.
These inspection activities constituted one complete sample as defined in
IP 71124.01-05.
b. Findings
No findings were identified.
.6 Radiation Worker Performance and Radiation Protection Technician Proficiency (02.07)
a. Inspection Scope
The inspectors observed radiation worker performance and assessed their performance
with respect to radiation protection work requirements, the level of radiological hazards
present, and RWP controls.
The inspectors assessed worker awareness of electronic alarming dosimeter set points,
stay times, or permissible dose for radiologically significant work as well as expected
response to alarms.
The inspectors observed radiation protection technician performance and assessed
whether the technicians were aware of the radiological conditions and RWP controls and
whether their performance was consistent with training and qualifications for the given
radiological hazards.
The inspectors observed radiation protection technician performance of radiation
surveys and assessed the appropriateness of the instruments being used, including
calibration and source checks.
These inspection activities constituted one complete sample as defined in
IP 71124.01-05.
b. Findings
No findings were identified.
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.7 Problem Identification and Resolution (02.08)
a. Inspection Scope
The inspectors assessed whether problems associated with radiological hazard
assessment and exposure controls were being identified at an appropriate threshold and
were properly addressed for resolution. For select problems, the inspectors assessed
the appropriateness of the corrective actions. The inspectors also assessed the
licensees program for reviewing and incorporating operating experience.
The inspectors reviewed select problems related to human performance errors and
assessed whether there was a similar cause and whether corrective actions taken
resolve the problems.
The inspectors reviewed select problems related to radiation protection technician error
and assessed whether there was a similar cause and whether corrective actions taken
resolve the problems.
These inspection activities constituted one complete sample as defined in
IP 71124.01-05.
b. Findings
No findings were identified.
2RS2 Occupational As-Low-As-Reasonably-Achievable Planning and Controls (71124.02)
.1 Radiological Work Planning (02.02)
a. Inspection Scope
The inspectors selected three to five work activities of the highest exposure significance
or involve work in high dose rate areas.
The inspectors reviewed the radiological work planning as-low-as-reasonably-achievable
(ALARA) evaluations, initial and revised exposure estimates, and exposure mitigation
requirements. The inspectors determined whether the licensee had reasonably grouped
the radiological work into work activities.
The inspectors assessed whether the licensees planning identified appropriate dose
reduction techniques; appropriately considered alternate reduction features; and defined
reasonable dose goals. The inspectors evaluated whether the licensees ALARA
assessment had taken into account decreased worker efficiency from use of respiratory
protective devices and/or heat stress mitigation equipment. The inspectors determined
whether the licensees work planning considered the use of remote technologies and
dose reduction insights from industry and plant-specific operating experience. The
inspectors assessed whether these ALARA requirements were integrated into work
procedures and/or RWP documents.
These inspection activities constituted a partial sample as defined in IP 71124.02-05.
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b. Findings
No findings were identified.
.2 Implementation of As-Low-As-Reasonably-Achievable and Radiological Work Controls
(02.04)
a. Inspection Scope
The inspectors reviewed the radiological administrative, operational, and engineering
controls planned for selected radiologically significant work activities and evaluated the
integration of these controls and ALARA requirements into work packages, work
procedures and/or RWPs.
The inspectors observed in-plant work activities and assessed whether the licensee had
effectively integrated the planned administrative, operational, and engineering controls
into the actual field work to maintain occupational exposure ALARA. The inspectors
observed pre-job briefings, and determined whether the planned controls were
discussed with workers. The inspectors evaluated the placement and use of shielding,
contamination controls, airborne controls, RWP controls, and other engineering work
controls against the ALARA plans.
The inspectors assessed licensee activities associated with work-in-progress to ensure
the licensee was tracking doses, performed timely in-progress reviews, and when jobs
did not trend as expected, appropriately communicated additional methods to be used to
reduce dose. The inspectors evaluated whether health physics and ALARA staff were
involved with the management of radiological work control when in-field activities
deviated from the planned controls. The inspectors assessed whether the Outage
Control Center and station management provided sufficient support for ALARA
re-planning.
The inspectors assessed the involvement of ALARA staff with emergent work activities
during maintenance and when possible, attended in-progress review discussions,
outage status meetings, and/or ALARA committee meetings.
These inspection activities constituted a partial sample as defined in IP 71124.02-05.
b. Findings
No findings were identified.
.3 Radiation Worker Performance (02.05)
a. Inspection Scope
The inspectors observed radiation worker and radiation protection technician
performance during work activities being performed in radiation areas, airborne
radioactivity areas, or HRAs to assess whether workers demonstrated the ALARA
philosophy in practice and followed procedures. The inspectors observed radiation
worker performance to evaluate whether the training and skill level was sufficient with
respect to the radiological hazards and the work involved.
21
The inspectors interviewed individuals from selected work groups to assess their
knowledge and awareness of planned and/or implemented radiological and ALARA work
controls.
These inspection activities constituted one complete sample as defined in
IP 71124.02-05.
b. Findings
No findings were identified.
.4 OTHER ACTIVITIES
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency
Preparedness, Public Radiation Safety, Occupational Radiation Safety, and
Security
4OA1 Performance Indicator Verification (71151)
.1 Unplanned Scrams per 7000 Critical Hours
a. Inspection Scope
The inspectors sampled licensee submittals for the Unplanned Scrams per 7000 Critical
Hours performance indicator (PI) for the period from the first quarter 2016 through the
fourth quarter 2016. To determine the accuracy of the PI data reported during those
periods, PI definitions and guidance contained in the Nuclear Energy Institute (NEI)
Document 99-02, Regulatory Assessment Performance Indicator Guideline,
Revision 7, dated August 31, 2013, were used. The inspectors reviewed the licensees
operator narrative logs, issue reports, event reports and NRC Integrated Inspection
Reports for the period of January 1, 2016 through December 31, 2016, to validate the
accuracy of the submittals. The inspectors also reviewed the licensees issue report
database to determine if any problems had been identified with the PI data collected or
transmitted for this indicator and none were identified. Documents reviewed are listed in
the Attachment to this report.
This inspection constituted one unplanned scram per 7000 critical hours sampled as
defined in IP 71151-05.
b. Findings
No findings were identified.
.2 Unplanned Scrams with Complications
a. Inspection Scope
The inspectors sampled licensee submittals for the Unplanned Scrams with
Complications PI for the period from the first quarter 2016 through the fourth
quarter 2016. To determine the accuracy of the PI data reported during those
periods, PI definitions and guidance contained in the NEI Document 99-02, Regulatory
Assessment Performance Indicator Guideline, Revision 7, dated August 31, 2013, were
used. The inspectors reviewed the licensees operator narrative logs, issue reports,
22
event reports and NRC Integrated Inspection Reports for the period of January 1, 2016
through December 31, 2016, to validate the accuracy of the submittals. The inspectors
also reviewed the licensees issue report database to determine if any problems had
been identified with the PI data collected or transmitted for this indicator and none were
identified. Documents reviewed are listed in the Attachment to this report.
This inspection constituted one unplanned scrams with complications sample as defined
in IP 71151-05.
b. Findings
No findings were identified.
.3 Unplanned Power Changes per 7000 Critical Hours
a. Inspection Scope
The inspectors sampled licensee submittals for the Unplanned Transients per
7000 Critical Hours PI for the period from the first quarter 2016 through the fourth
quarter 2016. To determine the accuracy of the PI data reported during those periods,
PI definitions and guidance contained in the NEI Document 99-02, Regulatory
Assessment Performance Indicator Guideline, Revision 7, dated August 31, 2013, were
used. The inspectors reviewed the licensees operator narrative logs, issue reports,
maintenance rule records, event reports and NRC Integrated Inspection Reports for the
period of January 1, 2016 through December 31, 2016, to validate the accuracy of the
submittals. The inspectors also reviewed the licensees issue report database to
determine if any problems had been identified with the PI data collected or transmitted
for this indicator and none were identified. Documents reviewed are listed in the
Attachment to this report.
This inspection constituted one unplanned transients per 7000 critical hours sample as
defined in IP 71151-05.
b. Findings
No findings were identified.
4OA2 Identification and Resolution of Problems (71152)
.1 Routine Review of Items Entered into the Corrective Action Program
a. Inspection Scope
As discussed in previous sections of this report, the inspectors routinely reviewed issues
during baseline inspection activities and plant status reviews to verify they were being
entered into the licensees CAP at an appropriate threshold, adequate attention was
being given to timely corrective actions, and adverse trends were identified and
addressed. Some minor issues were entered into the licensees CAP as a result of the
inspectors observations; however, they are not discussed in this report.
These routine reviews for the identification and resolution of problems did not constitute
any additional inspection samples. Instead, by procedure they were considered an
integral part of the inspections performed during the quarter.
23
b. Findings
No findings were identified.
.2 Semi-Annual Trend Review
a. Inspection Scope
The inspectors performed a review of the licensees CAP and associated documents
to identify trends that could indicate the existence of a more significant safety issue.
The inspectors review was focused on repetitive equipment issues, but also considered
the results of daily inspector CAP item screening discussed in Section 4OA2.1 above,
licensee trending efforts, and licensee human performance results. The inspectors
review nominally considered the 6-month period of July 1, 2016 through
December 31, 2016, although some examples expanded beyond those dates where the
scope of the trend warranted.
The review also included issues documented outside the CAP in major equipment
problem lists, repetitive and/or rework maintenance lists, departmental
problem/challenges lists, system health reports, quality assurance audit/surveillance
reports, self-assessment reports, and Maintenance Rule assessments. The inspectors
compared and contrasted their results with the results contained in the licensees
CAP trending reports. Corrective actions associated with a sample of the issues
identified in the licensees trending reports were reviewed for adequacy.
This review constituted one semi-annual trend review inspection sample as defined in
b. Findings
No findings were identified.
.3 Annual Follow-up of Selected Issues: Reviewed Licensee Corrective Actions for Failure
to Keep and Maintain Records that Include the Location and the Unique Identity of
Special Nuclear Material Items and Failure to Follow Written Material Control and
Accounting Procedures
a. Inspection Scope
During a review of items entered in the licensees CAP, the inspectors recognized
five CRs that focused on the licensees handling of Special Nuclear Material (SNM).
The first four CRs were written to document findings identified by the NRC during the
biennial Material Control & Accounting (MC&A) inspection, conducted in October and
November 2016. The fifth CR documented the movement of SNM from the nuclear
instrument cabinet in the fuel handling building to the instrument & calibration (I&C) hot
shop in the intermediate building without required documentation and independent
verification.
The inspectors reviewed the corrective actions in each of the CRs listed above. The
licensee has completed all corrective actions for CRs generated from the MC&A
inspection last year, with the exception of one. The corrective actions for the
undocumented transfer of SNM from its assigned storage location to the I&C hot shop
24
on intermediate building 654 elevation included documenting the return of the SNM to
the NI storage cabinet in the fuel handling building, a stand down on NI control and
accountability with the reactor engineering group, and human performance event
response. During their reviews of these five CRs, the inspectors made the following
observations.
- The four CRs written to address the NCV issued in Perry Nuclear Power
Plant - NRC Material Control and Accounting Program IR 05000440/2016406
were processed as Category-AF (adverse fix). Procedure NOP-LP-2001,
Corrective Action Program, Revision 38, states, in part, in Attachment 2,
Condition Report Evaluation Methods, that Fix - Evaluation Code F is not
sufficient for process, program, or equipment issues that result in: NRC
cited/non-cited violations. The inspectors documented this same observation in
Perry Nuclear Power PlantNRC Integrated IR 05000440/2016001 for CR
2015-11597, Potential NRC Violation concerning operation of the DG ventilation
fans, dated September 2, 2015, which was also documented as an NCV in Perry
Nuclear Power PlantNRC Integrated IR 05000440/2015003.
- NOP-LP-2001; Corrective Action Program, Revision 38, a quality procedure,
states, in part, that CRs shall be written to document receipt of NRC Findings or
Cited or Non-Cited Violations in accordance with NOBP-LP-4014 to specifically
address the issue(s) as stated in the wording received from the NRC, and include
actions to correct the finding or violation. Nuclear Operating Business Practice
NOBP-LP-4014; Managing Regulatory Interface; Revision 6, states, in part, in
Section 2.1.2, Adherence to this business practice is mandatory for NRC
inspections. NOBP-LP-4014, also states, in part, in Section 4.1.6.5, Ensure
separate CRs have been written for each potential and confirmed NRC inspector
finding or violation, recommending at least a causal evaluation both because of
the regulatory significance (violation of regulatory requirements) and to ensure
organizational factors contributing to cross-cutting aspects are considered. The
four CRs written for the MC&A NCV were categorized as adverse fix and did not
document or address the cross-cutting aspect, Change Management (H.3) in the
area of Human Performance. Additionally, no other CR was written to address
the cross-cutting aspect after the NCV was issued in Perry Nuclear Power
PlantNRC Material Control and Accounting Program IR 05000440/2016406.
The inspectors concluded that these were minor findings as there were only two
examples of the licensees failure to follow a quality procedure and there is no regulatory
requirement to write a condition report to address individual cross-cutting aspects
assigned to a NCV.
This review constituted one in-depth problem identification and resolution sample as
defined in IP 71152-05.
b. Findings
No findings were identified.
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.4 Annual Follow-up of Selected Issues: Review of Licensees Corrective Actions for
Failing to Perform an Engineering Evaluation for the Continued Functionality of the
Underdrain and Gravity Discharge Systems as a Result of Rock Salt Intrusion and its
Potential Long Term Corrosive Effects on the Systems Porous Concrete
a. Inspection Scope
In October 2016, the inspectors identified a severity level IV NCV of 10 CFR 50.59(d)(1),
Changes, Test, and Experiments, and associated finding for the licensees failure to
perform a written evaluation that provided the basis for the determination that a change
did not require a license amendment. Specifically, the licensee made a change pursuant
to 10 CFR 50.59(c) with the installation of grated manhole covers, replacing the rubber
gasket, watertight manhole covers for the underdrain and gravity discharge systems and
did not provide a basis for the determination that this change would not result in a more
than a minimal increase in the likelihood of occurrence of a malfunction of a system
structure or component important to safety. The licensee entered this issue into the
CAP as CR 2016-11864 and performed a prompt operability determination to show that
the underdrain and gravity drain systems remained functional while the engineering
change package was developed to support the change and bring the underdrain and
gravity discharge systems into compliance with the design basis. In January 2017 the
inspectors questioned the continued functionality of the underdrain and gravity discharge
systems based on the introduction of rock salt into the systems from the treatment of
roadways and travel paths during the winter months and its degrading effects on the
porous concrete in the systems. The licensee recognized that the prompt functionality
assessment did not address the effects of sodium chloride on the underdrain and gravity
discharge systems, nor its effects on the emergency service water system. The licensee
determined that continued functionality, in the short term, was reasonably assured as
degradation of the porous concrete was more of a long term concern and as such
continued functionality remained during the investigation and evaluation of the longer
term degradation of the porous concrete by the intrusion of rock salt into the systems.
The licensee evaluated the inspectors concerns and concluded that the introduction of
rock salt into the systems did not have any adverse impacts that would comprise the
expectation of continued functionality of the underdrain and gravity discharge systems,
the emergency service water system, or the plant buildings containing safety related
systems.
During their review of CR 2016-11864, the inspectors made the following observations.
- The licensees functionality assessment involving the removal of the solid
watertight gasketed covers, described in the licensees USAR and replacing
those covers with gratings in September 2016, did not take into account the
introduction of rock salt into the underdrain and gravity discharge systems, and
the intrusion of additional saline water into the emergency service water system.
- NOP-LP-2001; Corrective Action Program; Revision 38, in Attachment 1,
Adverse Condition or Non-Adverse Conditions, lists Condition that may result
in a NRC violation, or has significance within a regulatory context as an adverse
condition. The licensee failed to document the inspectors concerns, for a
potential condition adverse to quality in its CAP program from January 6, 2017
until March 13, 2017.
26
The inspectors concluded that these were minor findings as the first observation was a
violation that was determined to be minor because the failure to evaluate the intrusion of
rock salt did not impact the systems functionality and the second finding was only one
example of the licensees failure to follow a quality procedure.
This review constituted one in-depth problem identification and resolution sample as
defined in IP 71152-05.
b. Findings
No findings were identified.
4OA5 Other Activities
.1 (Closed) NRC Temporary Instruction 2515/192, Inspection of the Licensees Interim
Compensatory Measures Associated with the Open Phase Condition Design
Vulnerabilities in Electric Power Systems
a. Inspection Scope
The objective of this performance based temporary instruction (TI) is to verify
implementation of interim compensatory measures associated with an open phase
condition (OPC) design vulnerability in electric power system for operating reactors. The
inspectors conducted an inspection to determine if the licensee had implemented the
following interim compensatory measures. These compensatory measures are to
remain in place until permanent automatic detection and protection schemes are
installed and declared operable for OPC design vulnerability. The inspectors verified the
following:
- The licensee had identified and discussed with plant staff the lessons-learned
from the OPC events at the US operating plants including the Byron station OPC
event and its consequences. This includes conducting operator training for
promptly diagnosing, recognizing consequences, and responding to an OPC
event.
- The licensee had updated plant operating procedures to help operators promptly
diagnose and respond to OPC events on off-site power sources credited for safe
shutdown of the plant.
- The licensee had established and continue to implement periodic walkdown
activities to inspect switchyard equipment such as insulators, disconnect
switches, and transmission line and transformer connections associated with the
offsite power circuits to detect a visible OPC.
The licensee had ensured that routine maintenance and testing activities on switchyard
components have been implemented and maintained. As part of the maintenance and
testing activities, the licensee assessed and managed plant risk in accordance with 10
CFR 50.65(a) (4) requirements.
b. Findings and Observations
No findings of significance were identified.
27
.2 (Closed) Violation 05000440/2015010-01; Unqualified Radiation Protection Manager
On December 4, 2015, Notice of Violation 05000440/2015010-01 was issued for the
failure to take corrective action to comply with TS 5.3.1 and Regulatory Guide (RG) 1.8,
dated September 1975 regarding the qualifications of the individual performing duties of
the Radiation Protection Manager (RPM). On January 12, 2016, the licensee appointed
an individual as RPM with qualifications necessary to satisfy the requirements specified
in TS 5.3.1 and RG 1.8, dated September, 1975. The inspectors concluded the
licensees corrective actions were acceptable. This violation is closed.
.3 (Closed) Unresolved Item 05000440/2013008-03: Lack of Alternate Methods of Decay
Heat Removal
a. Inspection Scope
The NRC documented an unresolved item (URI) in Inspection Report 5000440/2013008
(ML13276A131) involving the unavailability of alternate methods of decay heat removal
that could be credited to combat a loss of SDC resulting from ESW inoperability and
while in MODE 4 with high decay heat load. The issue was left unresolved pending
further review and determination of NRC actions to resolve the issue. During this
inspection period, the inspectors consulted with the Office of Enforcement and Technical
Specification Branch of the Office of Nuclear Reactor Regulations about this issue.
The documents that were reviewed are included in the Attachment to this report. This
review did not represent an inspection sample. This URI is closed.
b. Findings
(1) Failure to Establish Procedures for Combating a Loss of Shutdown Cooling
Introduction: A finding of very-low safety significance and associated NCV of TS 5.4,
Procedures, was identified by the inspectors for the failure to implement procedures for
combating a loss of SDC. Specifically, the licensee did not have a procedure which
could be effectively implemented for combating a loss of SDC resulting from ESW
inoperability and during high decay heat load.
Description: As described in IR 05000440/2013008, the licensee was unable to meet
Technical Specification 3.4.10 regarding establishment of an alternate method for decay
heat removal on May 21, 2004, and on October 19, 2009, when one or both of the ESW
systems were declared inoperable during shutdown conditions. Specifically, TS Limiting
Condition of Operation (LCO) 3.4.10, Residual Heat Removal Shutdown Cooling
System - Cold Shutdown, requires, in part, two shutdown cooling subsystems operable
in MODE 4 when heat losses to the ambient were not sufficient to maintain average
reactor coolant temperature below 200 degrees Fahrenheit. With one or two shutdown
cooling subsystems inoperable, TS 3.4.10, Required Action A.1, required the licensee to
verify an alternate method of decay heat removal was available for each inoperable
shutdown cooling subsystem within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter. The
associated TS Basis described the alternate method as one that re-establishes backup
decay heat removal capabilities similar to the requirements of the LCO. In addition, it
stated, The required cooling capacity of the alternate method should be ensured by
verifying (by calculation or demonstration) its capability to maintain or reduce
temperature. As a result of these events, the licensee installed the Alternate Decay
28
Heat Removal (ADHR) system to aid in TS compliance, in MODE 4, by providing an
additional decay heat removal method that does not rely upon RHR or ESW. Similar
incidents also occurred on June 11, 2007, and on February 11, 2016. The 2004 and
2007 incidents resulted in NCVs, which were documented in IR 05000440/2004011 and
IR 05000440/2007005, respectively.
During the 2013 Evaluations of Changes, Tests and Experiments and Permanent Plant
Modifications Inspection (i.e., Mod/50.59 Inspection), the inspectors reviewed the
associated 10 CFR 50.59 evaluation (i.e., Evaluation 05-04712, Installation of ADHR
System) which stated, The intent of the ADHR system is to assure TS compliance in
MODE 4 by providing an additional alternate decay heat removal option that does not
rely upon RHR or ESW. However, the inspectors noted the following concerns:
- the ADHR system provided only one alternate method and its design was limited
to a heat removal rate which bounded the approximate core decay heat
production rate 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a plant shutdown from sustained 100 percent
power.
process flow to the heat exchanger was available. This condition was captured
in the CAP as CR 2013-06220. The associated functionality assessment
determined ADHR was limited to a heat removal rate which bounded the
approximate core decay heat production rate 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after a shutdown from
sustained 100 percent power for refueling outage 1R15.
- the licensee also revised Procedure ONI-E12-2, Loss of Decay Heat Removal,
by adding Attachment 11, Cold Shutdown Decay Heat Removal by Steaming.
This attachment contained instructions to establish one alternate method of
decay heat removal independent of ESW. However, its effectiveness to re-
establish backup decay heat removal capabilities similar to the requirements of
LCO 3.4.10 had not been verified. Specifically, the attachment included a note
stating, It will be necessary to validate the effectiveness of this attachment to
maintain or reduce reactor pressure vessel temperature (by Engineering
calculation or demonstration) if qualifying this as an ADHR method per TS 3.4.9
and 3.4.10. In response to the inspectors questions, the licensee estimated this
method was limited to a heat removal rate which bounded the approximate core
decay heat production rate 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after a shutdown from sustained 100
percent power.
- the ADHR system is maintained in a dry condition and requires more than an
hour to fill and vent in order to be declared operational.
Based on this information, the inspectors concluded that for the October 19, 2009,
occurrence, the plant failed to implement an alternate method of decay heat removal that
could be verified to be available within an hour following the loss of a train of ESW while
in Mode 4. The inspectors also noted that during normal shutdown conditions, the
licensee transitions from 100 percent power to MODE 4 in a few hours. For instance,
the transition to MODE 4 during the 1R13 refueling outage occurred in about five hours.
In the first three loss of SDC instances described above, the licensee submitted
Licensee Event Report (LERs) stating that the site could not demonstrate the
requirements of TS 3.4.10 and was operating in a condition prohibited by TS and
therefore reported the issue under 10 CFR 50.73(a)(2)(i)(B) as identified in
LER 4402004001, LER 4402007002, and LER 4402009003.
29
The licensee captured the inspectors concerns in their CAP as CR 2016-11987. The
corrective actions considered at the time of this inspection were to perform calculations
for various conditions to determine available alternatives for MODE 4 entry at less than
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or demonstrate alternatives are viable; and provide procedural guidance based
on the results.
Analysis: The inspectors determined the failure to implement procedures for combating
a loss of SDC resulting from all applicable conditions was contrary to TS 5.4,
Procedures, and was a performance deficiency. The inspectors determined the
performance deficiency was more-than-minor because it was associated with the
Mitigating Systems cornerstone attribute of design control 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
licensee cannot verify alternate methods of decay heat removal are available, as
required by Required Action A.1 upon discovery that LCO 3.4.10, is not met.
The inspectors determined the finding could be evaluated using the Significance
Determination Process in accordance with Inspection Manual Chapter 0609,
Significance Determination Process, Attachment 4, Initial Characterization of
Findings, dated October 7, 2016 and Appendix G, Shutdown Operations Significance
Determination Process, Exhibit 3, Mitigating Systems Screening Questions, dated
May 9, 2014. The finding screened as very-low safety significance (Green) because it
did not affect the operability or Probabilistic Risk Assessment functionality of any
mitigating SSCs.
The inspectors 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.
Enforcement: Technical Specification 5.4, Procedures, stated, in part, that written
procedures shall be established, implemented, and maintained covering the applicable
procedures recommended in Regulatory Guide 1.33, Quality Assurance Program
Requirements, Revision 2, Appendix A. Regulatory Guide 1.33, Appendix A, Section 6,
addressed Procedures for Combating Emergencies and Other Significant Events, and
sub-section 6.h, addressed Loss of Shutdown Cooling. In addition, TS LCO 3.0.2
requires that upon discovery of a failure to meet an LCO, the Required Actions of the
associated Conditions shall be met. With one or two SDC subsystems inoperable,
Required Action A.1 of TS 3.4.10 requires the licensee to verify an alternate method of
decay heat removal was available for each inoperable SDC subsystem within one hour
and once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter. Alternate methods of decay heat removal that satisfy
this TS requirement are defined in the associated TS Basis as those that re-establish
backup decay heat removal capabilities similar to the requirements of TS 3.4.10.
Contrary to the above, on October 9, 2009, the licensee failed to implement a procedure
recommended in Appendix A of Regulatory Guide 1.33, Revision 2. Specifically, the
licensee could not implement its procedure for combating a loss of SDC resulting from
ESW inoperability with high decay heat load. As a result, upon discovery of a failure to
meet LCO 3.4.10 during a loss of ESW with high decay heat load, Required Action A.1
could not be met as required by LCO 3.0.2. On October 19, 2009, at 0429 hours0.00497 days <br />0.119 hours <br />7.093254e-4 weeks <br />1.632345e-4 months <br /> the
train B of SDC was declared inoperable, as a result of the loss of train B of ESW, and
the licensee was unable to implement procedure ONI-E12-2 because an alternate
30
method of decay heat removal with a capability similar to the requirements of TS 3.4.10
could not be verified to be available within one hour.
At the time of this inspection period, the licensee was still evaluating its planned
corrective actions. However, the inspectors determined that the continued
non-compliance did not present an immediate safety concern because all shutdown
cooling subsystems were expected to be operable if needed during this inspection
period.
Because this violation was of very-low safety significance and was entered into the
licensees CAP as CR 2016-11931, this violation is being treated as a NCV, consistent
with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000440/2017001-01, Failure
to Implement Procedures for Combating a Loss of Shutdown Cooling).
4OA6 Management Meetings
.1 Exit Meeting Summary
On April 13, 2017, the inspectors presented the inspection results to Mr. F. Payne and
other members of the licensee staff. The licensee acknowledged the issues presented.
The inspectors confirmed that none of the potential report input discussed was
considered proprietary.
.2 Interim Exit Meetings
- On March 10, 2017, an Interim exit meeting was conducted for the inspection
results of the ISI review with Mr. D. Hamilton and other members of the licensee
staff. The inspectors confirmed that none of the potential report input discussed
was considered proprietary.
- On March 24, 2017, an Interim exit meeting was conducted for the inspection
results of the Radiation Safety Program review with Mr. D. Hamilton and other
members of the licensee staff. The inspectors confirmed that none of the
potential report input discussed was considered proprietary.
- On April 21, 2017, an Interim exit meeting was conducted for the inspection
results for the closure of URI 05000440/2013008-03 to Mr. D. Hamilton, and
other members of the licensee staff. The licensee acknowledged the issues
presented. The inspectors confirmed that none of the potential report input
discussed was considered proprietary.
4OA7 Licensee-Identified Violations
The following violation of very low significance (Green) was identified by the licensee
and is a violation of NRC requirements, which meet the criteria of the NRC Enforcement
Policy for being dispositioned as a Non-Cited Violation (NCV).
In part, 10 CFR 20.1703 (c)(5) states, The licensee shall implement and maintain a
respiratory protection program that includesDetermination by a physician that the
individual user is medically fit to use respiratory protection equipment.
Contrary to the above, the licensee identified that an individual wore a powered air
purifying respirator (PAPR) three times during the period of March 6-7, 2017 for the
purpose of radiological protection without the required medical determination. This was
31
entered into the licensees corrective action program, CR 2017-02957, Vessel
Technician Wore PAPR Three Times without Being Qualified. The significance of this
violation was determined in accordance with IMC 0609 Appendix C, Occupational
Radiation Safety Significance Determination Process dated August 19, 2008. This
violation was determined to be of very low safety significance (Green), because this
violation was not associated with ALARA Planning or Work Controls, there was no
overexposure nor substantial potential for overexposure and the ability to access dose
was not compromised.
ATTACHMENT: SUPPLEMENTAL INFORMATION
32
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
D. Hamilton, Site Vice-President
F. Payne, General Plant Manager
D. Saltz, Performance Improvement Director
J. Ellis, Maintenance Director
D. Reeves, Site Engineering Director
L. Zerr, Regulatory Compliance
D. Lieb, Technical Services Supervisor
J. Truxall, Inspection Response Team
S. Lee, Health Physicist
J. Spahr, RPM
U.S. Nuclear Regulatory Commission
J. Cameron, Chief, Reactor Projects Branch 4
D. Hills, Chief, Engineering Branch 1
H. Peterson, Chief, Health Physics and Incident Response Branch
M. Jeffers, Chief, Engineering Branch 2
Attachment
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
05000440/2017001-01 NCV Failure to Implement Procedures for Combating a
Loss of Shutdown Cooling
Closed
TI 2515/192 TI Inspection of the Licensees Interim Compensatory
Measures Associated with the Open Phase Condition
Design Vulnerabilities in Electrical Power System
05000440/2015010-01 NOV Unqualified Radiation Protection Manager
05000440/2013008-03 URI Lack of Alternate Methods of Decay Heat Removal
05000440/2017001-01 NCV Failure to Implement Procedures for Combating a
Loss of Shutdown Cooling
Discussed
None.
2
LIST OF DOCUMENTS REVIEWED
The following is a partial list of documents reviewed during the inspection. Inclusion on this list
does not imply that the NRC inspector reviewed the documents in their entirety, but rather that
selected sections or 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.
1R04 Equipment Alignment
- VLI-C41; Standby Liquid Control System Valve Lineup Instruction; Revision 8
- SOI-R43; Division 1 and 2 Diesel Generator System; Revision 45
- VLI-R44; Division 1 and 2 Diesel Generator Starting Air System; Revision 6
- VLI-R45; Division 1 and 2 Diesel Generator Fuel Oil System (Unit 1); Revision 5
- VLI-R48; Division 1 and 2 Diesel Generator Exhaust, Intake and Crankcase Systems;
Revision 6
- VLI-P45; Emergency Service Water System; Revision 19
- Dwg 302-0351-00000; Standby Diesel Generator Starting Air; Revision GG
- Dwg 302-0352-00000; Standby Diesel Generator Fuel Oil System; Revision LL
- Dwg 302-0353-00000; Standby Diesel Generator Lube Oil; Revision T
- Dwg 302-0354-00000; Standby Diesel Generator Jacket Water; Revision V
- Dwg 302-0357-00000; Div 1 and Div 2 Diesel Air Dryer Diagrams; Revision H
- VLI-M25/26; Control Room HVAC and Emergency Recirculation System; Revision 7
- Drawing (Dwg) 912-0610-00000; Control Room HVAC and Emergency Recirculation System;
Revision GG
- ELI-S11; Power Transformer; Revision 9
- VLI-E22A; High Pressure Core Spray; Revision 10
- SVI-E22-T2001; HPCS Pump and Valve Operability Test; Revision 28
- SOI-E22A; High Pressure Core Spray System; Revision 36
- NOP-OP-1001; Clearance/Tagging Program; Revision 24
- CR 2016-14542; 60 dpm Weld Leak Upstream of 1E22F514, HPCS CST Test Press Inst Root;
dated December 22, 2016
- VLI-R47/E22B; Division 3 Diesel Generator Lube Oil System (Unit 1); Revision 4
- VLI-R48/E22B; Division 3 Diesel Generator Exhaust, Intake and Crankcase Systems;
Revision 1
- VLI-R46/E22B; Division 3 Diesel Generator Jacket Water System; Revision 6
- VLI-R45/E22B; Division 3 Diesel Generator Fuel Oil System (Unit 1); Revision 3
- VLI-R44/E22B; Division 3 Diesel Generator Starting Air System; Revision 10
1R05 Fire Protection
- FPI-1DG; Diesel Generator Building, Revision 8;
- CR 2017-01686; Unplanned Fire Impairment for DG-108 Fire Door; dated February 15, 2017
- SOI-M43; Diesel Generator Building Ventilation System; Revision 15;
- FPI-1RB; Reactor Building; Revision 4;
- CR 2017-02172; Post Event Critique for ONI-P54 Entry; dated February 28, 2017
- FPI-0FH; Fuel Handling Building; Revision 5;
- Dwg 023-0012-00000; USAR Drawing; Fire Protection Evaluation; Intermediate Building and
Fuel Handling Building Plan; EL. 620-6" Revision J
- FPI-0CC; Control Complex; Revision 10
- ONI-P54; Fire; Revision 21
3
- FPI-A-B02; Fire Drill Critique; dated February 2, 2017
- FPI-A-B02; Fire Drill Planning Guide; dated February 2, 2017
- FPI-A-B02; Fire Brigade Drills; dated February 2, 2017
- FPI-A-B02; Fire Drill Critique; dated February 27, 2017
- FPI-A-B02; Fire Drill Planning Guide; dated February 27, 2017
- FPI-A-B02; Fire Brigade Drills; dated February 27, 2017
- Triple Tech, Inc. Fire Protection Expert; Fire Report; dated February 27, 2017
- FPI-A-B02; Fire Drill Critique; dated February 6, 2017
- FPI-A-B02; Fire Drill Planning Guide; dated February 6, 2017
- FPI-A-B02; Fire Brigade Drills; dated February 6, 2017
1R08 Inservice Inspection
- CR 2017-02668; Workers Could not Locate Component for Examination Resulting in
Additional Dose; March 10, 2017
- CR 2017-02666; NRC ID: NQI-1042 Contains Unnecessary Requirements; March 10, 2017
- CR 2017-02683; NRC Inspector Question on Evaluation of Indication Found During 1R15;
March 10, 2017
- CR 2015-04803; Indication Identified in H9 Shroud Support Plate to Reactor Vessel Wall
Weld; May 7, 2015
- CR 2015-03374; Snubber 1E12-H0765 has Gap; March 14, 2015
- CR 2015-02652;High Pressure Core Spray Variable Spring Hanger 1E22-H0071 has a
Potential Relevant Condition; March 2, 2015
- CR 2016-01423; Deficient Welds Identified During Extent of Condition for CR 2016-01071 on
1B33F0013A an 1B33F0014A; January 30, 2016
- CR 2015-011884; the Response to CR 2015-04064 is not Technically Correct;
September 9, 2015
- CR 2015-05471; Unsatisfactory Non Destructive Examination Results on VT-3 Examination of
1P45-H0703, Work Package 20592665; April 19, 2015
- CR 2015-05245; FME: Foreign Material Found During Core Verification; April 15, 2015
- CR 2015-04098; FME: Paper Dropped in the Fuel Pool During Reactor Maintenance;
March 26, 2015
- CR 2015-09189; Target Rock Solenoid Valve 10 CFR Part 21 Report for Defect of Soft-Seated
Solenoid Operated Valve components; July 26, 2016
- CR 2015-09596; No Hardware Disposition Performed for Out of Tolerance Snubber;
July 15, 2015
- CR 2015-15320; VT-2 Exam Marked N/A when ASME Parts were Replaced;
November 9, 2015
- NQI-0942; Magnetic Particle Examination; Revision 20
- NQI-1042; Visual Examination; Revision 18
- NOP-CC-5762; Appendix VIII Procedure for Ultrasonic Examination of Ferritic Pipe Welds;
Revision 2
- NOP-CC-5765; Appendix VIII Procedure for Straight Beam Ultrasonic Examination of Bolts
and Studs; Revision 4
- NOP-CC-5767; Appendix VIII Procedure: Site Demonstration Protocol for Ultrasonic Bolting
Examination; Revision 0
- UT-17-E004; 12 Valve F053A to Pipe; March 9, 2017
- UT-17-E005; RPV Closure Head Studs; March 9, 2017
- UT-17-E003; 6 Valve F019 to Pipe; Dated March 3, 2017
- 942-17A-001; MT of Piping Support Welded Attachment 1E21-H0020-WA; March 1, 2017
- 1042-17-023; VT-3 of 1E21-H0025; February 21, 2017
4
- 1042-17-083; VT-3 of 1E21-H0004; March 9, 2017
- 1042-17-028; VT-3 of 1E12-H0633; February 24, 2017
1R11 Licensed Operator Requalification Program
- Cycle 1 2017 Evaluated Scenario C2; OTLC - 3058201701_PY - SGC2; Revision 0
- IOI-0003; Power Changes; Revision 65
- IOI-0004; Shutdown; Revision 23
- IOI-0008; Shutdown by Manual Reactor Scram; Revision 8
1R12 Maintenance Effectiveness
- Perry Nuclear Power Plant, Plant Health Report 2016-02 - R22 - Metal Clad switch Gear
(15 KV and 5KV); dated February 2, 2012
- CR 2016-02048; Loss of EH11 Divisional Bus Results in a Loss of Shutdown Cooling; dated
February 11, 2016
- NORM-ER-3107; FENOC Power Fuses; Revision 02; WO
- Perry Nuclear Power Plant, Plant Health Report 2016-02 - C51 - Neutron Monitoring; dated
February 2, 2012
- CR 2015-09050; Digital Indication for IRM G Range does not Indicate Properly; dated
July 4, 2015
- CR 2016-13146; IRM F did not Track Properly when Inserting for Approximately 15 Seconds;
dated November 4, 2016
- WO 200574346; Replace IRM C/G Range Switch Assembly in Panel 1H3P0680; dated
March 7, 2017
1R13 Maintenance Risk Assessments and Emergent Work Control
- Perry Work Implementation Schedule; Week 04, Period 7, Division 3, Forecast On-Line
Probabilistic Risk Assessment January 9, 2017 to January 15, 2017; Revision 1
- CR 2016-14542; 60 dpm Weld Leak Upstream of 1E22F514, HPCS CST Test Press Inst Root;
dated December 22, 2016
- NOP-OP-1007; Risk Management; Revision 23
- NOBP-OP-0012; Operator Work-Arounds, Burdens and Control Room Deficiencies and
Operations Aggregate Assessment;
- NOPL-AD-0010; Integrated Risk Management; Revision 1
- PDB-C0011; PSA Pre-Solved Configurations for On-Line Risk; Revision 8
- PYBP-POS-2-2; Protected Equipment Postings; Revision 12
- 1R16 Shutdown Defense-in-Depth Report; Revision 1
- eSOMS Plant Narrative Log; dated March 1, 2017
- eSOMS Plant Narrative Log; dated March 3, 2017
1R15 Operability Determinations and Functionality Assessments
- eSOMS Plant Narrative Logs; dated January 3, 2017
- eSOMS Plant Narrative Logs; dated January 6 and 7, 2017
- CR 2017-00066; SLC Pump B Out of Service Alarm - Squib Continuity; dated
January 3, 2017
- EER 601079696; Ability to Function of Standby Liquid Control; dated January 5, 2017
- CR 2017-00787; RCIC Watertight Found Open; dated January 24, 2017
- OAI-0201; Operations General Instructions and Operating Practices; Revision 43
5
- CR 2017-02974; NRC ID: Response to CR 2017-00787 did not Address USAR Bases;
March 16, 2017
- CR 2016-11864; NRC ID: Underdrain Manhole Covers Changed to Grating vs Watertight
Covers; October 4, 2016
- CR 2017-02787; NRC ID: Concern with Continued Functionality of the Underdrain System;
dated March 13, 2017
- Calculation P72-006 Addendum 1; Evaluation of Chemical-Deicing Agents Being Introduced
into the Underdrain System (P72) at the Perry Nuclear Power Plant; Revision 0; dated
March 17, 2017
- eSOMS Plant Narrative Logs; dated March 17, 2017
- eSOMS Plant Narrative Logs; dated March 20, 2017
- PMI-0113; Plant Underdrain System Maintenance and Inspection; Revision 4
- NOP-OP-1009; Operability Determinations and Functionality Assessments; Revision 6
- CR 2016-12837; PA-PY-16-005: Plant Underdrain System Plot Plan Containds Inaccurate
Information; dated October 27, 2016
- CR 2016-14283; Unapproved Deviations from Engineering Change Packages Result in
Challenges to Ongoing Flood Hazards Analyses; dated December 14, 2016
- CR 2017-01054; 10 CFR 50.59 not Completed for CAN 13-0802-006, Door Barriers; dated
January 31, 2017
- CR 2017-02864; Degraded Mechanical Snubber 1E12-H0211
- WO 200646510; RHR Snubber Removal and Instillation
- Functional Test Data Sheet; Report No: FT-17-0076
1R18 Plant Modifications
- ECP 16-03476-000; DG Ventilation Bypass Switch Modification; dated January 19, 2017;
Revision 1
- ECP 16-03476-001; Diesel Generator Supply Fan 1M43C0001A/2A CO2 Trip Override Switch
Modification; Revision 0
- ECP 16-03476-001; Diesel Generator Supply Fan 1M43C0001A/2A CO2 Trip Override Switch
Modification; Revision 1
- ECP 16-03476-002; Diesel Generator Supply Fan 1M43C0001A/2A CO2 Trip Override Switch
Modification; Revision 0
- ECP 16-03476-001; Diesel Generator Supply Fan 1M43C0001A/2A CO2 Trip Override Switch
Modification; Revision 1
- ONI-P54; Revision 21
- SOI-M43; Diesel Generator Building Ventilation System; Revision 15
- PTI-P54-P0034A; Division 1 Diesel Generator CO2 Systems Detection and Operability Test;
Revision 9
- PTI-P54-P0034B; Division 2 Diesel Generator CO2 Systems Detection/Operability Test;
Revision 8
- WO 200692307; Implement ECP 16-0178-001 for Division 1 DG Vent Replaces M3-S7 and
Removes M43-K34; dated January 28, 2017
- WO 200692420; Implement ECP 16-0178-002 for Division 2 DG Vent Replaces M43-S8 and
Removes M43-K35; dated February 1, 2017
1R19 Post-Maintenance Testing
- Perry Nuclear Power Plant Plan of Action for Operations Challenge; Standby Liquid Control
Out of Service Alarm; dated January 3, 2017
6
- CR 2017-00066; SLC pump B Out of Service Alarm - Squib Continuity; dated
January 3, 2017
- WO 200702893; Loss of Power Squib Vlv Continuity; dated January 11, 2017
- Notification 601079731; Standby Liquid Control B Squib Continuity Valve; dated
January 3, 2017
- SOI-E51 Section 7.17; Controlled Startup from Standby Readiness to CST Mode Using
Remote Shutdown Panel Controller; Revision 34
- PTI-E51-P0003; RCIC Terry Turbine Overspeed Trip Test; Revision 10
- WO 200518407; replace DC type M relay in ED1A09-C; dated January 25, 2017
- WO 200518408; replace DC type M relay in ED1A09-E; dated January 25, 2017
- WO 200518405; replace DC type M relay in ED1A09-P; dated January 25, 2017
- WO 200592472; Replace ESW Pump B PY-1P45C0001B per ECP 14-0082; dated
March 26, 2017
- PTI-P45-P0002; ESW System Loop B Flow and Differential Pressure Test; dated
March 26, 2017
- PTI-P45-P0002; ESW System Loop B Flow and Differential Pressure Test; dated
March 27, 2017
- SVI-P45-T2002; ESW Pump B and Valve Operability Test; dated March 27, 2017
- WO 200643216; ECP 15-0257-001: Install HICO XFMR; dated March 27, 2017
- WO 200643314; ECP 15-0057-005: Open Phase Mod RFO16; dated March 27, 2017
- CR 2017-03060; Unit 2 Transformer Install Order did not Follow Vendor Install Guidance;
dated March 17, 2017
- WO 200692307; Implement ECP 16-0178-001 for Division 1 DG Ventilation Replaces M43-S7
and Removes M43-K34; dated January 28, 2017
- WO 200692420; Implement ECP 16-0178-002 for Division 2 DG Ventilation Replaces M43-S8
and Removes M43-K35; dated February 1, 2017
- CR 2017-03381; Division 2 DG Fan Damper Indication (M43-F071B); dated March 24, 2017
1R20 Outage Activities
- eSOMS Plant Narrative Logs; dated March 4, 2017
- eSOMS Plant Narrative Logs; dated March 5, 2017
- eSOMS Plant Narrative Logs; dated March 8, 2017
- NOBP-OM-2003; Outage Control Center Guidelines; Revision 9
- IOI-3; Power Changes; Revision 64
- GEN-MNT-0002; Generation Rigging and Lifting Manual; Revision 1
- NOP-WM-5003; Rigging, Lifting and Load Handling; Revision 5
- GEN-SAF-0001; Generation Personal Safety Manual; Revision 2
- GMI-0185B; Reactor Vessel Assembly; Revision 13
- GMI-0226; Refuel Floor Maintenance Activities; Revision 5
- IOI-0020; Operations with the Potential to Drain the Reactor Vessel; Revision 0
- IOI-0001; Cold Startup; Revision 44
- FTI-A0009; Estimated Range for Critical; Revision 07
- Reactivity Plan; Startup 127; Part 2
- Reactivity Plan; Beginning of Cycle 17; Startup 127; Part 1
- NOBP-OP-1004-02 Revision 00; Evolution Specific Reactivity Plan
- NOBP-OM-4010; Restart Readiness For Plant Outages; Revision 04
- NOBP-OM-4010; Restart Readiness For Plant Outages
- NOBP-OM-4010-06 System Engineer Readiness Affirmation; Revision 00
- NRC Integrated Inspection Report 05000440/2016004 and 05000440/2016501
7
- Enforcement Guidance Memorandum 11-003; Dispositioning Boiling Water Reactor Licensee
Noncompliance with Technical Specification Containment Requirements During Operations
with a Potential for Draining the Reactor Vessel; Revision 1; dated December 20, 2012
- EGM 11-003; Dispositioning Boiling Water Reactor Licensee Noncompliance with Technical
Specification Containment Requirements During OPDRV; Revision 3, dated January 15, 2016
- Outage Preparation Two Month Review Meeting; dated January 6, 2017
- Hope Creek Generating Station Unit 1; LER 2012-003
- Monticello Nuclear Generating Plant NRC Integrated and Power Update Inspection Report and
Exercise of Enforcement Discretion 05000263/2015002
- Restart Readiness Meeting Package PY-1R17; dated March 29, 2017
- CR 2016-12326; Temporary Valve Left Installed on the HICO Start-up Transformer on Load
Tap Changer (OLTC) Compartment; dated October 14, 2016
- CR 2017-02562; PA-PY-17-01 Walkdown Level Indication Protection Scheme; dated
March 8, 2017
- CR 2017-02620; Protected Equipment Posting Found Outside of Normal Posting Position;
dated March 9, 2017
- CR 2017-02748; FME: Foreign Material Entered the Upper Pools, Northwest Corner; dated
March 12, 2017
- CR 2017-02816; 1R16 LLRT: Check Valve 1P51-F530 Failed to Pressurize during
SVI-P51-T9308; dated March 13, 2017
- CR 2017-02713; Unexpected Isolation during SVI-B21-T1402 Logic Functional; dated
March 11, 2017
- CR 2017-02755; 1R16 Trending: Potential Trend in Risk Recognition and Preparation; dated
March 12, 2017
- CR 2017-02819; Relief Valve 1C41F0029A Failed its Bench Test; dated March 13, 2017
- CR 2017-02822; FME Floating in Reactor Cavity; dated March 13, 2017
- CR 2017-02844;1N11F0045B Found to Mechanically Bind when Cycling in the Close Direction
during Troubleshooting; dated March 14, 2017
- CR 2017-02858; USAR Discrepancies Related to EDG Non-Critical Trips; dated
March 14, 2017
- CR 2017-02876; Two Long Term Rod Control System Items Missed in 1R16 Tracked by
PLCOs Since 2015; dated March 14, 2017
- CR 2017-02900; Snubber E12H0280 Failed Drag Test; dated March 15, 2017
- CR 2017-02980; 1R16 LLRTs SVI-P53-T9305 and SVI-P53-T9312 Partial Performance; dated
March 16, 2017
- CR 2017-03004; 1B33F0067B would not Close Remotely; dated March 17, 2017
- CR 2017-03046; 1R16 LLRT: MSIV Outboard Accumulator Check Valves 1B21-F029B and
1B21-F029C Exceed 849.4 sccm; dated March 17, 2017
- CR 2017-03076; DW EDS Pump A Discharge Failed to Open in AUTO or Manual with Signal
and Power Indication Present; dated March 18, 2017
- CR 2017-03103; Post Event Critique for EOP-03 Entry; dated March 19, 2017
- CR 2017-03109; Loss of FME in Dryer Pool while Cutting Dry Tube; dated March 19, 2017
- CR 2017-03110; Start-up Transformer - HICO Drawing and Valve Location Nameplate
Contain Information not Consistent with Transformer; dated March 19, 2017
- CR 2017-03121; Bus D-1-B Ground Alarm, dated March 19, 2017
- CR 2017-03127; LPCS and RHR A Operated on Minimum Flow for Greater than One Hour;
dated March 20, 2017
- CR 2017-03127; Request Engineering to Evaluate the Effects on Both Pumps.
CA06-11480-01
- CR 2017-03128; C85 (Steam Bypass & Pressure Regulation) Buffer Checking Circuit Test
Card Failed calibration; dated March 20, 2017
8
- CR 2017-03129; Access to Verify Site Qualifications; dated March 20, 2017
- CR 2017-03131; Access to the CR/Correction Action Program at Perry Nuclear Power Plant;
dated March 20, 2017
- CR 2017-03142; Violation of NOBP-LP-1113; dated April 19, 2017
- CR 2017-03159; Document of FM in Eye; dated March 20, 2017
- CR 2017-03264; Emergency Service Water (ESW) B Pump Discharge Head Mounting
Flange Corrosion; dated March 22, 2017
- CR 2017-03269; as Found Condition of 1C11F0160B; dated March 22, 2017
- CR 2017-03281; Ultrasound Thickness Results were Less than Minimum Wall Thickness on
the Condenser Water Box C and D Common Vent Lines; dated March 22, 2017
- CR 2017-03360; NRC Identified PAP-0114 Needs Clarification; dated March 24, 2017
1R22 Surveillance Testing
- SVI-P45-T2003; HPCS ESW Pump and Valve Operability Test; dated January 9, 2017
- PDB-R0002; Perry Surveillance Test Interval List; Revision 1
- NOP-ER-3030; Surveillance Frequency Control Program; Revision 0
- NOP-WM-2003; Work Management Surveillance Program; Revision 8
- SVI-C51-T0050-G; OPRM Channel G Calibration for 1C51-K603G; Revision 9
- ABB OPRM Report; January 26, 2017
- SVI-C51-T5001-G; OPRM Channel G Functional for 1C51-K603G; Revision 7
- SVI-B33-T025-B; EOC-RPT Breaker ARC Suppression Response Time for 1B33A-CB4A and
1B33A-CB4B; dated March 4, 2017
- FTI-F0031; Volumetrics and FENOC Leak Rate Monitors Testing Instruction; Revision 4
- SVI-C41-T2001-B; Standby Liquid Control B Pump and Valve Operability Test;
February 2, 2017
- SVI-R43-T1317; Diesel Generator Start and Load Division 1; Revision 19
1EP6 Drill Evaluation
- Cycle 1 2017 Evaluated Scenario C2; OTLC - 3058201701_PY - SGC2; Revision 0
2RS1 Radiological Hazard Assessment and Exposure Controls
- Self-Assessment Radiation Protection Program Reviews 2013-2015; October 20, 2016
- Self-Assessment Pre NRC Assessment of Rad Hazards, ALARA and Airborne, March 3, 2017
- HPI-D0001; Radiation and Contamination Survey Techniques; Revision 25
- HPI-L0009; Discrete Particle Control; Revision 6
- IOI-0017; Drywell Entry and Access Control; Revision 23
- NOP-OP-4101; Access Controls for Radiologically Controlled Areas; Revision 12
- NOP-OP-4102; Air Sampling; Revision 5
- NOP-OP-4107; Radiation Work Permit (RWP); Revision 16
- NOP-OP-4502; Control of Radioactive Material; Revision 4
- NOP-OP-4701; Radiological Survey Documentation; Revision 1
- NOBP-OP-4009; Radworker Expectations; Revision 6
- RWP 176018; 1R16 Reactor Disassembly Activities; Revisions 0-2
- RWP 176070; 1R16 1G33 RWCU HX Valve Replacement Activities; Revision 1
- RWP 176048; 1R16 Undervessel Activities; Revisions 0-1
- R16 Administrative Dose Extension Authorizations; Various Records
- Periodic Barrier/Barricade Surveillance; March 17, 2017
- SVI-E31-T5190; Sealed Source Leak Test & Inventory, Revision 7
- Radioactive Source Inventory and Leak Test; February 1, 2017
9
- National Source Tracking System 2017 Annual Inventory Reconciliation; January 23, 2017
- National Source Tracking System 2017 Annual Inventory Reconciliation; March 9, 2017
- Electronic Dosimeter Alarm Records; Various March 2017 Records
- Radiological Air Samples; Various Records
- Radiological Surveys; Various Records
- CR 2016-14249; PCM Alarm Due to Radionuclide Medical Treatment; December 13, 2016
- CR 2017-02770; Emerging Trend in Radiological Performance; March 12, 2017
- CR 2017-02869; Unbriefed Dose Rate Alarm in RWCU Heat Exchanger Room;
March 14, 2017
- CR 2017-02881; Radiological Posting Deficiencies; March 14, 2017
- CR 2017-02920; Unbriefed Dose Rate Alarm Condenser Bay 600; March 15, 2017
- CR 2017-02957; Vessel Technician Wore PAPR Three Times Without Being Qualified;
March 16, 2017
- CR 2017-03052; Drywell Entry; March 17, 2017
- CR 2017-03067; Level 2 PCE 17-01; March 18, 2017
- CR 2017-03360; NRC Identified PAP-0114 Needs Clarification; March 24, 2017
2RS2 Occupational ALARA Planning and Controls
- NOP-OP-4005; ALARA Program; Revision 6
- Station ALARA Committee Meeting Information; March 18, 2017
- Station ALARA Committee Meeting Information; March 21, 2017
- ALARA Work In Progress Reviews; Various Documents
- ALARA Pre-Planning Work Sheet; Various Documents
- ALARA Brief Checklist; Various Documents
- RWCU Project Overview Documentation; Undated
- Radiological Source Term Cycle 16; October 21, 2016
- HIS-20 RWP Summary Report; March 21, 2017
- 16 Refuel Outage Dose Estimate Trend; March 23, 2017
- Reactor Water Co-60 Outage Comparison; 1R9 through 1R16
4OA1 Performance Indicator Verification
- NOBP-LP-4012; NRC Performance Indicators; Revision 5
- NOBP-LP-4012-01; Unplanned Scrams per 7,000 Critical Hours; Revision 2; January 2016
through December 2016
- NOBP-LP-4012-02; Unplanned Scrams with Complications (USwC); Revision 3; January 2016
through December 2016
- NOBP-LP-4012-03; Unplanned Power Changes per 7,000 Critical Hours; Revision 2;
January 2016 through December 2016
4OA2 Problem Identification and Resolution
- CR 2017-01845; Inclined Fuel Transfer Winch Cable and Cable Guide Issues; dated
February 20, 2017
- CR 2017-01041; LTC Back up Controller Removed from Unit 2 Transformer for Tap Changer
Failed Testing; dated January 31, 2017
- CR 2016-13090; MC&A Inspection: Independent Verification of Special Nuclear Material
Movement; dated November 3, 2016
- CR 2016-08737; NRC ID: Control Complex 638 Elevation Fire Barrier Wall Non-Compliance;
dated July 13, 2016
10
- CR 2017-02787; NRC ID: Concern with Continued Functionality of the Underdrain System;
dated March 13, 2017
- CR 2017-01754; Trend CR - Under Drain Manhole Repeat Failures due to Calcium Build Up;
dated February 17, 2017
- NOP-NF-3001; Perry: Nuclear Instrumentation Movement Checklist; Revision 9; Order
Number 200664435; dated February 13, 2017
- NOP-NF-3002; Special Nuclear Material Physical Inventory; Revision 2
- CR 2017-01492; LPRM Issued to Hotshop without Move Sheet; dated February 9, 2017
- CR 2016-08737; NRC ID: Control Complex 638 Elevation Fire Barrier Wall Non-Compliance;
dated July 13, 2016
- CR 2016-09239; MS-C-16-07-16: Finding - Flooding in Level B Material Storage Areas in the
Perry Warehouses; dated July 27, 2016
- CR 2016-09746; 2016 NRC FLEX Inspection: PM Development for FLEX Communications
System; dated August 10, 2016
- CR 2016-09965; Incomplete Information Regarding Respirator Mask Inspections Presented to
NRC during Security Inspection on August 11, 2016; dated August 18, 2016
- CR 2016-10301; Inadvertent Misposition of Plant Equipment during Weekly Routine D17 Filter
Change Outs; dated August 29, 2016
- CR 2016-11251; R35 Cathodic Protection Two Test Wells Found not Meeting PTI-R35-P0002
Acceptance Criteria; dated September 23, 2016
- CR 2016-11373; Reactor Feed Booster Pump A Tripped; dated September 25, 2016
- CR 2016-11864; NRC ID: Underdrain Manhole Covers Changed to Grating vs. Watertight
Covers; dated October 4, 2016
- CR 2016-12485; Electric Duct Heater Controllers Installed without Appropriate Design
Documentation; dated October 18, 2016
- CR 2016-12755; Planned Work Activities Assigned to Incorrect Unit 1 Start-up Transformer
Work Window; dated October 25, 2016
- CR 2016-12935; Emergency Closed Cooling System Valve Found Out of Position; dated
October 31, 2016
- CR 2016-13183; During Performance of PTI-R43-P0006B CB-1, CB-2, CB-3, CB-4 DC
Breakers Trip when Loaded; dated November 6, 2016
- CR 2016-13267; NRC ID Cyber Security PI&R Inspection: the Patching of Kiosks does not
Meet the Intent of NEI 0809, Appendix D 1.19; dated November 9, 2016
- CR 2016-13418; During the Processing of Control Rod Blades (CRB) - CRB 0010 Dropped to
the Cask Pit Floor; dated November 15, 2016
- CR 2016-13541; MS-C-16-11-24: Finding: Perry Emergency Plan is not in compliance with
10CFR50 Appendix E for Training Descriptions
- CR 2016-14141; CR-2015-16646, Configuration Control, Actions Determined to be Ineffective;
dated December 9, 2016
- CR 2016-14445; QC Concerns Found during Plant Walk Down; dated December 19, 2016
- CR 2016-14627; Loss of Power SLC B Squib Vlv Continuity during SLC A Pump and Valve
Testings
4OA5 Other Activities
- SVI-R10-T5228; On-Site Power Distribution System Verification; Revision 7
- SVI-R10-T5227; Off-Site Power Availability Verification; Revision 8
- Notification 600883101; Outside Rounds: Change Switchyard House and switchyard Breaker
Checks to Weekly on Sunday; dated March 1, 2014
- CR 2012-11563; Actions Needed as a Result of Investigation of Event Report 12-14 Level 2;
dated July 25, 2012
11
- CR 2013-17631; Design Vulnerability in Perry Electrical Power System - Open Phase
Condition not Detectable Under Certain Plant Loading Conditions; dated November 1, 2013
- CR 2017-02344; NRC ID: Open Phase Operator Training for Industry Operating Experience
at Bryon Station; dated March 4, 2017
- CR 2017-01171; the Unit 1 Open Phase Protection System (OPPS) Panel Indicated an
Injection Abnormal Alarm; dated February 2, 2017
- CR 2017-01016; Fleet Open Phase Protection System - HVAC Unit Failures; dated
January 30, 2017
- CR 2009-66216; Unable to Meet Tech Spec Action Statement Due to ESW B Inoperability;
October 19, 2009
- Evaluation 05-04712; Installation of Alternate Decay Heat Removal System;
December 21, 2012
- ONI-E12-2; Loss of Decay Heat Removal; Revision 31
- CR 2016-11987; 2016 NRC Triennial Heat Sink Inspection: Proposed Non-Cited Violation
Related to URI 2013008; October 10, 2016
- SOI0-G40 (ADHR); Alternate Decay Heat Removal; Revision 3
- PAP-1925; Shutdown Defense in Depth Assessment and Management; Revision 18
12
LIST OF ACRONYMS USED
ADAMS Agencywide Document Access Management System
ADHR Alternate Decay Heat Removal
ALARA As Low as Reasonably Achievable
ASME American Society of Mechanical Engineers
CAP Corrective Action Program
CFR Code of Federal Regulations
CR Condition Report
DG Diesel Generator
EDG Emergency Diesel Generator
ESW Emergency Service Water
HVAC Heating, Ventilation, and Air Conditioning
I&C Instrument and Calibration
IMC Inspection Manual Chapter
IP Inspection Procedure
IR Inspection Report
ISI Inservice Inspection
LCO Limiting Condition of Operation
LER Licensee Event Report
MCA Material Control and Accounting
NCV Non-Cited Violation
NEI Nuclear Energy Institute
NOBP Nuclear Operating Business Practice
NRC Nuclear Regulatory Commission
OPDRV Operations with Potential of Draining the Reactor Vessel
OPRM Osculating Power Range Monitor
OSP Outage Safety Plan
PI Performance Indicator
PM Post Maintenance
RCA Radiologically Controlled Area
RCIC Reactor Core Isolation Cooling
RFO Refueling Outage
RWP Radiation Work Permits
SSC Structure, System, and Component
TS Technical Specification
UFSAR Updated Final Safety Analysis Report
URI Unresolved Item
USAR Updated Safety Analysis Report
UT Ultrasonic Examination
VT-3 Visual-3 Examination