ML11213A091
| ML11213A091 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 07/29/2011 |
| From: | Ring M NRC/RGN-III/DRP/B1 |
| To: | Pacilio M Exelon Generation Co, Exelon Nuclear |
| References | |
| IR-11-003 | |
| Download: ML11213A091 (62) | |
See also: IR 05000461/2011003
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION III
2443 WARRENVILLE ROAD, SUITE 210
LISLE, IL 60532-4352
July 29, 2011
Mr. Michael J. Pacilio
Senior Vice President, Exelon Generation Company, LLC
President and Chief Nuclear Officer (CNO), Exelon Nuclear
4300 Winfield Road
Warrenville, IL 60555
SUBJECT: CLINTON POWER STATION, NRC INTEGRATED INSPECTION REPORT
Dear Mr. Pacilio:
On June 30, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at
your Clinton Power Station. The enclosed report documents the inspection results, which were
discussed on July 13, 2011, with Mr. W. Noll and other members of your staff.
This inspection examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
Based on the results of this inspection, two NRC-identified findings of very low safety
significance were identified. Both of these findings were determined to involve violations of
NRC requirements. Additionally, one licensee-identified violation, which was determined to be
of very low safety significance, was reviewed by the inspectors and is listed in this report.
Because of the very low safety significance and because they were entered into your
corrective action program, the NRC is treating the above inspector-identified and
licensee-identified violations as non-cited violations (NCVs) consistent with Section VI.A.1 of
the NRC Enforcement Policy. If you contest any NCV, you should provide a response within
30 days of the date of this inspection report, with the basis for your denial to the 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,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident
Inspector at Clinton Power Station. In addition, if you disagree with the cross-cutting aspect
assigned to any finding in this report in this report, you should provide a response within
30 days of the date of this inspection report, with the basis for your disagreement to the
Regional Administrator, Region III, and the NRC Resident Inspector at Clinton Power Station.
The information you provide will be considered in accordance with Inspection Manual
Chapter 0305.
M. Pacilio -2-
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter,
its enclosure, and your response (if any) will be available electronically for public inspection in
the NRC Public Document Room or from the Publicly Available Records System (PARS)
component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Website
at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Mark A. Ring, Chief
Branch 1
Division of Reactor Projects
Docket No. 50-461
License No. NPF-62
Enclosure: Inspection Report 05000461/2011-003
w/Attachment: Supplemental Information
cc w/encl: Distribution via ListServ
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No: 50-461
License No: NPF-62
Report No: 05000461/2011-003
Licensee: Exelon Generation Company, LLC
Facility: Clinton Power Station, Unit 1
Location: Clinton, IL
Dates: April 1 through June 30, 2011
Inspectors: B. Kemker, Senior Resident Inspector
D. Lords, Resident Inspector
C. Brown, Reactor Inspector
J. Cassidy, Senior Health Physicist
A. Dunlop, Senior Reactor Engineer
M. Jones Jr., Reactor Inspector
R. Winter, Reactor Inspector
S. Mischke, Resident Inspector, Illinois Emergency
Management Agency
Approved by: M. Ring, Chief
Branch 1
Division of Reactor Projects
Enclosure
TABLE OF CONTENTS
SUMMARY OF FINDINGS ......................................................................................................... 1
REPORT DETAILS .................................................................................................................... 3
Summary of Plant Status ........................................................................................................ 3
1. REACTOR SAFETY..................................................................................................... 3
1R01 Adverse Weather Protection (71111.01) ........................................................... 3
1R04 Equipment Alignment (71111.04)...................................................................... 5
1R05 Fire Protection (71111.05) ................................................................................ 6
1R06 Flooding Protection Measures (71111.06) ........................................................ 7
1R07 Heat Sink Performance (71111.07)................................................................... 8
1R11 Licensed Operator Requalification Program (71111.11) ..................................11
1R12 Maintenance Effectiveness (71111.12) ............................................................11
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13) .......12
1R15 Operability Evaluations (71111.15) ..................................................................13
1R18 Plant Modifications (71111.18) ........................................................................14
1R19 Post-Maintenance Testing (71111.19) .............................................................14
1R22 Surveillance Testing (71111.22) ......................................................................15
1EP6 Drill Evaluation (71114.06) ..............................................................................21
2. RADIATION SAFETY ..................................................................................................22
2RS3 In-Plant Airborne Radioactivity Control and Mitigation (71124.03) ...................22
2RS4 Occupational Dose Assessment (71124.04) ....................................................26
4. OTHER ACTIVITIES ...................................................................................................30
4OA1 Performance Indicator Verification (71151) ......................................................30
4OA2 Identification and Resolution of Problems (71152) ...........................................31
4OA3 Followup of Events and Notices of Enforcement Discretion (71153) ................32
4OA5 Other Activities ................................................................................................33
4OA6 Management Meetings ....................................................................................37
4OA7 Licensee-Identified Violations ..........................................................................38
SUPPLEMENTAL INFORMATION............................................................................................. 1
KEY POINTS OF CONTACT .................................................................................................. 1
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED........................................................ 2
LIST OF DOCUMENTS REVIEWED ...................................................................................... 4
LIST OF ACRONYMS USED.................................................................................................17
Enclosure
SUMMARY OF FINDINGS
IR 05000461/2011-003, 04/01/11 - 06/30/11, Clinton Power Station, Unit 1, Heat Sink
Performance, Surveillance Testing.
This report covers a three-month period of inspection by the resident inspectors and announced
baseline inspections by regional inspectors. Two Green findings, both of which had an
associated non-cited violation, were identified. The significance of most findings is indicated by
their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609,
Significance Determination Process (SDP). Findings for which the SDP does not apply may
be Green or be assigned a severity level after NRC management review. The NRCs program
for overseeing the safe operation of commercial nuclear power reactors is described in
NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.
A. NRC-Identified and Self-Revealed Findings
Cornerstone: Mitigating Systems
- Green. The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B,
Criterion III, Design Control, having very low safety significance for the failure to
include all of the applicable heat loads in the Reactor Core Isolation Cooling (RCIC)
Room heat up calculation and not having a calculation of record for the RCIC Room heat
up under a station blackout (SBO) scenario. The licensee entered this issue into the
corrective action program and performed preliminary calculations to verify that the issues
did not exceed any design limits.
The performance deficiency was determined to be more than minor because it
was associated with the Mitigating Systems Cornerstone attribute of Equipment
Performance, and affected the cornerstone objective of ensuring the availability,
reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences. The finding screened as very low safety significance
because the licensee determined the RCIC Room cooler was capable of removing the
additional heat load; and RCIC Room temperature remained within the design limits
without the room cooler during a SBO scenario. The inspectors determined that this
finding did not represent current licensee performance and no cross-cutting aspect was
assigned. (Section 1R07.1.b.(1))
Cornerstone: Initiating Events
- Green. The inspectors identified a finding of very low safety significance (Green) with an
associated non-cited violation of Technical Specification Surveillance Requirement
(TSSR) 3.4.6.1. The licensee failed to correctly incorporate the required test pressure
limits of the TSSR into the surveillance test procedure and subsequently tested multiple
reactor coolant system (RCS) pressure isolation valves (PIVs) at pressures greater than
the maximum test pressure of 1025 pounds per square inch gauge, invalidating the
testing. The licensee performed a risk assessment of the missed surveillance in
accordance with TSSR 3.0.3, which determined that completion of the surveillance could
be delayed up to the 24-month surveillance interval without a significant increase in plant
risk. The licensee also completed an operability evaluation for the TS nonconformance
and concluded that there was reasonable assurance that the affected RCS PIVs were
operable based on engineering judgment.
1 Enclosure
The finding was of more than minor significance because it affected the Initiating Events
Cornerstone and was associated with the Procedure Quality attribute. Specifically, the
licensee did not correctly incorporate the required test pressure limits of TSSR 3.4.6.1
into the surveillance test procedure. This resulted in testing multiple RCS PIVs at
pressures greater than the maximum test pressure of 1025 psig. The finding was
determined to be a licensee performance deficiency of very low safety significance
because the finding would not result in exceeding the TS limit for RCS leakage and
would not have likely affected mitigation systems resulting in a loss of safety function.
The inspectors concluded that because the licensees missed opportunity to correct the
test pressure discrepancy in its surveillance test procedure occurred in January 2005
and no other more recent opportunities reasonably existed to identify and correct the
problem, this issue would not be reflective of current licensee performance and no
cross-cutting aspect was identified. (Section 1R22.b.(1))
B. Licensee-Identified Violations
A violation of very low safety significance that was identified by the licensee has been
reviewed by the inspectors. Corrective actions planned or taken by the licensee have
been entered into the licensees corrective action program. The violation and corrective
action tracking numbers are listed in Section 4OA7 of this report.
2 Enclosure
REPORT DETAILS
Summary of Plant Status
The unit was operated at or near full power during the inspection period with the following
exceptions:
On April 2, 2011, the licensee reduced power to about 48 percent (%) to perform repairs on a
main condenser tube leak. The unit was returned to full power the following day.
On April 8, 2011, the licensee reduced power to about 82% to perform control rod pattern
adjustments. The unit was returned to full power the same day.
On May 22, 2011, the licensee reduced power to about 80% to perform control rod sequence
exchange, scram time testing and recovery of two control rods following hydraulic control unit
maintenance, control rod settle testing, and main turbine control/stop/intermediate valve and
main steam isolation valve testing. The unit was returned to full power the same day.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection (71111.01)
.1 Readiness For Impending Hot Summer Weather Conditions
a. Inspection Scope
The inspectors evaluated the licensees preparations for hot summer weather conditions,
focusing on the electrical distribution system and the plant chilled water system.
During the weeks of May 23, 2011, and June 20, 2011, the inspectors performed a
detailed review of severe weather and plant de-winterization procedures and performed
general area plant walkdowns. The inspectors focused on plant-specific design features
and implementation of procedures for responding to or mitigating the effects of hot
summer weather conditions on the operation of the plant. The inspectors reviewed
system health reports and system engineering summer readiness review documents for
the above systems.
Additionally, the inspectors verified that adverse weather related issues were entered
into the licensees corrective action program with the appropriate characterization and
significance. Selected action requests were reviewed to verify that corrective actions
were appropriate and implemented as scheduled.
This inspection constituted one seasonal extreme weather readiness inspection sample
as defined in Inspection Procedure (IP) 71111.01.
b. Findings
No findings were identified.
3 Enclosure
.2 Summer Readiness of Offsite and Alternate Alternating Current (AC) Power Systems
a. Inspection Scope
The inspectors evaluated the licensees plant features and procedures for operation and
continued availability of offsite and alternate AC power systems. The inspectors
interviewed plant personnel and reviewed the licensees communications protocols
between the Transmission System Operator (TSO) and the plant to verify that the
appropriate information was being exchanged when issues arose that could impact the
offsite power system. Aspects considered in the inspectors review included:
- The actions to be taken when notified by the TSO that the post-trip voltage of the
offsite power system at the plant will not be acceptable to assure the continued
operation of the safety related loads without transferring to the onsite power
supply;
- The compensatory actions identified to be performed if it is not possible to predict
the post-trip voltage at the plant for the current grid conditions;
- The required re-assessment of plant risk based on maintenance activities that
could affect grid reliability, or the ability of the transmission system to provide
offsite power; and
- The required communications between the plant and the TSO when changes at
the plant could impact the transmission system, or when the capability of the
transmission system to provide adequate offsite power is challenged.
The inspectors performed a walkdown of the switchyard with a plant maintenance
engineer to observe the material condition of the offsite power sources. The inspectors
also reviewed the status of outstanding work orders to assess whether corrective actions
for any degraded conditions were scheduled with the TSO with the appropriate priority.
This inspection constituted one offsite and alternate AC power systems readiness
inspection sample as defined in IP 71111.01.
b. Findings
No findings were identified.
.3 Readiness For Impending Adverse Weather Condition - Tornado/High Winds
a. Inspection Scope
Since thunderstorms with potential tornados and high winds were forecast in the vicinity
of the facility for the week of April 18, 2011, the inspectors reviewed the licensees
overall preparations/protection for the expected conditions. The inspectors toured the
plant grounds in the vicinity of the main power transformers, unit auxiliary transformer,
reserve auxiliary transformers, emergency reserve auxiliary transformer, and static volt
amp reactive compensators to look for loose debris, which if present could become
missiles during a tornado or with high winds. During the inspections, the inspectors
focused on plant-specific design features and the licensees procedure used to respond
to tornado and high winds conditions.
4 Enclosure
This inspection constituted one readiness for impending adverse weather condition
inspection sample as defined in IP 71111.01.
b. Findings
No findings were identified.
.4 Readiness to Cope with External Flooding
a. Inspection Scope
The inspectors reviewed flood protection barriers and procedures for coping with
external flooding at the plant. The Clinton Power Station has limited susceptibility to
external flooding as described in Section 3.4.1.1 of the Updated Final Safety Analysis
Report (UFSAR) and Section 5.2 of the Individual Plant Examination for External Events
Report. The inspectors reviewed CPS 4303.02, Abnormal Lake Level, Revision 10, to
assess the adequacy of the licensee response to external flooding conditions.
The inspectors conducted a walkdown of the Lake Screen House, including the
shutdown service water pump rooms. The inspectors assessed the condition of water
tight door seals; the sealing of equipment floor plugs, electrical conduits, holes or
penetrations in floors and walls between the pump rooms; and the condition of room
floor drains, sumps, and sump pumps.
Additionally, the inspectors verified that external flooding protection issues were entered
into the licensees corrective action program with the appropriate characterization and
significance. Selected action requests were reviewed to verify that corrective actions
were appropriate and implemented as scheduled.
This inspection constituted one external flooding readiness inspection sample as defined
in IP 71111.01.
b. Findings
No findings were identified.
1R04 Equipment Alignment (71111.04)
.1 Quarterly Partial System Walkdowns (71111.04Q)
a. Inspection Scope
The inspectors performed partial system walkdowns of the following risk-significant
systems:
- Standby Gas Treatment (VG) System Train B during planned maintenance on
VG System Train A;
- Control Room Ventilation (VC) System Train B during planned maintenance on
VC System Train A; and
- AC Power Distribution System (selected portions of risk-significant system).
5 Enclosure
The inspectors selected these systems based on their risk significance relative to the
Reactor Safety Cornerstones. The inspectors reviewed operating procedures, system
diagrams, Technical Specification (TS) requirements, and the impact of ongoing work
activities on redundant trains of equipment. The inspectors verified that conditions did
not exist 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 were aligned correctly and available as necessary.
In addition, the inspectors verified that equipment alignment problems were entered into
the licensees corrective action program with the appropriate characterization and
significance. Selected action requests were reviewed to verify that corrective actions
were appropriate and implemented as scheduled.
This inspection constituted three partial system walkdown inspection samples as defined
in IP 71111.04.
b. Findings
No findings were identified.
1R05 Fire Protection (71111.05)
.1 Routine Resident Inspector Tours (71111.05Q)
a. Inspection Scope
The inspectors performed fire protection tours in the following plant areas:
- Fire Zone A-1e, General Access Area (West) - Elevation 7370;
- Fire Zone R-1j, Dry Active Waste Baler Room - Elevation 737'0";
- Fire Zone R-1n, Paint and Oil Storage Room - Elevation 737'0"; and
- Fire Zone T-1k, General Access Area (West) - Elevation 781'0".
The inspectors verified that transient combustibles and ignition sources were
appropriately controlled and assessed the material condition of fire suppression
systems, manual fire fighting equipment, smoke detection systems, fire barriers and
emergency lighting units. 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; that the licensees fire plan was in alignment with actual conditions; and that fire
doors, dampers, and penetration seals appeared to be in satisfactory condition.
In addition, the inspectors verified that fire protection related problems were entered into
the licensees corrective action program with the appropriate characterization and
significance. Selected action requests were reviewed to verify that corrective actions
were appropriate and implemented as scheduled.
This inspection constituted four quarterly fire protection inspection samples as defined in
IP 71111.05AQ.
6 Enclosure
b. Findings
No findings were identified.
.2 Fire Protection - Drill Observation (71111.05A)
a. Inspection Scope
During an announced drill on May 26, 2011, associated with a simulated fire in the
Condensate Booster Pump Room, the inspectors assessed the timeliness of the
fire brigade in arriving at the scene, the fire fighting equipment brought to the scene,
the donning of fire protective clothing, the effectiveness of communications, and the
exercise of command and control by the fire brigade leader. The inspectors also
assessed the acceptance criteria for the drill objectives; the rigor and thoroughness of
the post-drill critique; and verified that fire protection drill issues were being entered into
the licensee's corrective action program with the appropriate characterization and
significance.
This inspection constituted one annual fire protection drill inspection sample as defined
in IP 71111.05AQ.
b. Findings
No findings were identified.
1R06 Flooding Protection Measures (71111.06)
a. Inspection Scope
The inspectors reviewed selected risk important plant design features and licensee
procedures intended to protect the plant and its safety-related equipment from internal
flooding events. The inspectors reviewed flood analyses and design documents,
including the UFSAR, engineering calculations, and abnormal operating procedures to
identify licensee commitments. In addition, the inspectors reviewed licensee drawings to
identify areas and equipment that may be affected by internal flooding caused by the
failure or misalignment of nearby sources of water, such as the fire suppression or the
service water systems. The inspectors also reviewed the licensees corrective action
documents with respect to past flood-related items identified in the corrective action
program to verify the adequacy of the corrective actions. The inspectors performed a
walkdown of the following plant areas to assess the adequacy of watertight doors and
verify drains and sumps were clear of debris and were operable, and that the licensee
complied with its commitments:
- Shutdown Service Water Pump Rooms; and
- Turbine Building Basement - Elevation 7020.
This inspection constituted one internal flooding inspection sample as defined in
7 Enclosure
b. Findings
No findings were identified.
1R07 Heat Sink Performance (71111.07)
.1 Annual Heat Sink Performance (71111.07A)
a. Inspection Scope
The inspectors reviewed the licensees maintenance activities for the Division 2
Inverter Room cooler (1VX13SB). Specifically, the review included the program for
testing and analysis of the room cooler, which was cleaned, inspected, and evaluated.
The inspectors assessed the as-found and as-left condition of the heat exchanger by
direct observation and document reviews to verify that no deficiencies existed that would
adversely impact the heat exchangers ability to transfer heat to the shutdown service
water system and to ensure that the licensee was adequately addressing problems that
could affect the performance of the heat exchanger. The inspectors observed portions
of inspection and cleaning activities, and reviewed documentation to verify that the
inspection acceptance criteria specified in procedure ER-AA-340-1002, Service Water
Heat Exchanger Inspection Guide, Revision 4, were satisfactorily met.
This inspection constituted one annual heat sink inspection sample as defined in
b. Findings
No findings were identified.
.2 Triennial Review of Heat Sink Performance (71111.07T)
a. Inspection Scope
The inspectors reviewed operability determinations, completed surveillances, vendor
manual information, associated calculations, performance test results and cooler
inspection results associated with the Reactor Core Isolation Cooling (RCIC) Room
cooler and Control Room chillers. These heat exchangers/coolers were chosen based
on their risk significance in the licensees probabilistic safety analysis, their important
safety-related mitigating system support functions, their operating history, and their
relatively low margin.
For the RCIC Room cooler and the Control Room chillers, the inspectors reviewed the
methods and results of heat exchanger performance inspections. The inspectors
verified the methods used to inspect and clean heat exchangers were consistent with
as-found conditions identified and expected degradation trends and industry standards,
the licensees inspection and cleaning activities had established acceptance criteria
consistent with industry standards, and the as-found results were recorded, evaluated,
and appropriately dispositioned such that the as-left condition was acceptable.
In addition, the inspectors verified the condition and operation of the RCIC Room cooler
and the Control Room chillers were consistent with design assumptions in heat transfer
calculations and as described in the UFSAR. This included verification that the number
8 Enclosure
of plugged tubes was within pre-established limits based on capacity and heat transfer
assumptions. The inspectors verified the licensee evaluated the potential for water
hammer and established adequate controls and operational limits to prevent heat
exchanger degradation due to excessive flow-induced vibration during operation.
In addition, eddy current test reports and visual inspection records were reviewed to
determine the structural integrity of the heat exchanger.
The inspectors also witnessed the inspection of the RCIC Room cooler to look for
indications of macrofouling that includes live or dead mussels and clams, plant
material, or silt.
The inspectors verified the performance of the ultimate heat sink (UHS) and
safety-related shutdown service water (SX) system and their subcomponents, such as
piping, intake screens, pumps, valves, etc., by tests or other equivalent methods to
ensure availability and accessibility to the in-plant cooling water systems.
The inspectors reviewed completed surveillances, associated calculations, buried pipe
inspection results, chemistry monitoring program, sedimentation monitoring procedures,
condition reports, and work orders to ensure the condition of the UHS and the
SX system.
The inspectors also verified pipe stress analyses, direct and indirect buried pipe
inspection test results, pump vibration data, and trends associated with the SX system to
ensure that acceptance criteria were being satisfied and the as-found inspection results
were recorded, evaluated, and appropriately dispositioned, such that the as left condition
was acceptable.
The inspectors also conducted walkdowns of the service water intake structure and the
SX Pump Rooms to verify the general condition of the system and associated
subsystems.
b. Findings
(1) Deficiencies with RCIC Room Heat Up Analyses
Introduction
A finding of very low safety significance and associated non-cited violation of 10 CFR 50,
Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure
to include all of the applicable heat loads in the RCIC Room heat up calculation under
loss-of-coolant-accident (LOCA) and not having a calculation of record for the RCIC
room heat up under a station blackout (SBO) scenario.
Description
The inspectors reviewed calculation VY-01, VY System Cooling Load Calculation,
to verify the RCIC Room cooler was capable of removing heat generated in the RCIC
Room under various scenarios and that the room temperature would remain within the
design limit of 180 degrees Fahrenheit (°F). The inspectors identified that the RCIC
water leg pump was not listed as one of the rooms heat sources for several scenarios,
including LOCA and shutdown conditions. Although the heat load associated with the
pumps motor was small, under LOCA conditions, the available calculated margin was
only 4.6%.
9 Enclosure
In addition, the inspectors requested the analysis of the RCIC Room under an SBO
scenario. The licensee determined that this analysis had been inadvertently deleted
from calculation 3C10-1088-001, Revision 4, SBO Coping Assessment. As such there
was no calculation of record to address this scenario.
The licensee initiated action request (AR) 01206227 to address these concerns.
Based on the 5 horsepower RCIC water leg pump motor, the licensee determined there
was a 3% increase in heat load for the room, which reduced the available margin to
approximately 1.7%. During the licensees review, they also identified two
conservatisms in the calculation where pipe temperatures were assumed to be higher
than the temperatures that would be experienced during the scenario. Removal of these
conservatisms could increase the available margin to approximately 10%. With respect
to the SBO scenario, the licensee used the methodology in the calculation inputting
conservative room heat up loads and verified that the room temperature would be 157°F,
which was below the 180°F limit.
The inspectors concluded that, based on these evaluations, the RCIC Room
temperature would remain within the required limits during the various scenarios.
Analysis
The inspectors determined that the failure to include all heat loads in the RCIC Room
heat up calculations and to have a calculation of record for the RCIC Room heat up
under an SBO scenario was a performance deficiency. The performance deficiency was
determined to be more than minor because it was associated with the Mitigating
Systems Cornerstone attribute of Equipment Performance, and affected the cornerstone
objective of ensuring the availability, reliability, and capability of systems that respond to
initiating events to prevent undesirable consequences. Specifically, the RCIC Room
heat up calculation did not include the RCIC water leg pump motor, which would have
added an additional 3% heat load reducing the available margin to 1.7%. In addition,
no calculation of record existed for the RCIC Room heat up under an SBO scenario to
verify the room would remain within the design temperature limits.
The inspectors determined the finding could be evaluated using the Significance
Determination Process (SDP) in accordance with IMC 0609, Significance Determination
Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of
Findings, Table 4a for the Mitigating System Cornerstone. The finding screened as
very low safety significance (Green) because the finding was not a design or
qualification deficiency, did not represent a loss of system safety function, and did not
screen as potentially risk significant due to a seismic, flooding, or severe weather
initiating event. In addition, the licensee performed preliminary calculations to verify that
the RCIC Room cooler was capable of removing the additional heat load; and the RCIC
Room temperature remained within the design limits without the room cooler during a
SBO scenario.
Cross-Cutting Aspects
The inspectors determined there was no cross-cutting aspect associated with this finding
because this was a legacy design issue and, therefore, was not reflective of current
performance.
10 Enclosure
Enforcement
10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that measures
shall be established to assure that applicable regulatory requirements and the design
basis are correctly translated into specifications, drawings, procedures, and instructions.
Contrary to the above, as of April 22, 2011, the licensee did not correctly translate
applicable regulatory requirements and the design basis into specifications and
procedures. Specifically, the RCIC Room heat up calculation did not include the RCIC
water leg pump motor heat load and there was no analysis for the RCIC Room heat up
under an SBO scenario. Because this violation was of very low safety significance and it
was entered into the licensees corrective action program as AR 01206227, this violation
is being treated as an non-cited violation consistent with Section 2.3.2 of the NRC
Enforcement Policy (NCV 05000461/2011003-01, Deficiencies with RCIC Room
Heat Up Analyses).
1R11 Licensed Operator Requalification Program (71111.11)
.1 Resident Inspector Quarterly Review (71111.11Q)
a. Inspection Scope
The inspectors observed licensed operators during simulator training on June 22, 2011.
The inspectors assessed the operators response to the simulated events focusing on
alarm response, command and control of crew activities, communication practices,
procedural adherence, and implementation of Emergency Plan requirements.
The inspectors also observed the post-training critique to assess the ability of licensee
evaluators and operating crews to self-identify performance deficiencies. The crews
performance in these areas was compared to pre-established operator action
expectations and successful critical task completion requirements.
This inspection constituted one quarterly licensed operator requalification inspection
sample as defined in IP 71111.11.
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness (71111.12)
a. Inspection Scope
The inspectors evaluated the licensee's handling of selected degraded performance
issues involving the following risk-significant structures, systems, and components
(SSCs):
- Radiation Monitoring System.
The inspectors assessed performance issues with respect to the reliability, availability,
and condition monitoring of the SSCs. Specifically, the inspectors independently verified
the licensee's handling of SSC performance or condition problems in terms of:
11 Enclosure
- Appropriate work practices;
- Identifying and addressing common cause failures;
- Scoping of SSCs in accordance with 10 CFR 50.65(b);
- Characterizing SSC reliability issues;
- Tracking SSC unavailability;
- Trending key parameters (condition monitoring);
- 10 CFR 50.65(a)(1) or (a)(2) classification and reclassification; and
- Appropriateness of performance criteria for SSC functions classified (a)(2) and/or
appropriateness and adequacy of goals and corrective actions for SSC functions
classified (a)(1).
In addition, the inspectors verified that problems associated with the effectiveness of
plant maintenance were entered into the licensee's corrective action program with the
appropriate characterization and significance. Selected action requests were reviewed
to verify that corrective actions were appropriate and implemented as scheduled.
This inspection constituted one maintenance effectiveness inspection sample as defined
in IP 71111.12.
b. Findings
No findings were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)
a. Inspection Scope
The inspectors reviewed the licensee's evaluation and management of plant risk for
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:
- Planned maintenance during the week of April 11-15 on the Division 2 Diesel
- Planned maintenance during the week of May 2-6 on the RCIC System;
- Planned maintenance during the week of May 31-June 3 on the Division 1
Automatic Depressurization System, Division 1 VG System, Removal of Control
Building Tornado Missile Barrier, and Division 1 Essential Switchgear Heat
Removal System; and
- Emergent maintenance during week of June 13-17 to address steam leaks in the
Turbine Building Heater Bay.
These activities were selected based on their potential risk significance relative to
the Reactor Safety Cornerstones. As applicable for each of the above activities, the
inspectors reviewed the scope of maintenance work in the plants daily schedule,
reviewed Control Room logs, verified that plant risk assessments were completed as
required by 10 CFR 50.65(a)(4) prior to commencing maintenance activities, discussed
the results of the assessment with the licensees Probabilistic Risk Analyst and/or Shift
Technical Advisor, and verified that plant conditions were consistent with the risk
assessment assumptions. The inspectors also reviewed TS requirements and walked
12 Enclosure
down portions of redundant safety systems, when applicable, to verify that risk analysis
assumptions were valid, that redundant safety related plant equipment necessary to
minimize risk was available for use, and that applicable requirements were met.
In addition, the inspectors verified that maintenance risk related problems were
entered into the licensees corrective action program with the appropriate significance
characterization. Selected action requests were reviewed to verify that corrective
actions were appropriate and implemented as scheduled.
This inspection constituted four maintenance risk assessment inspection samples as
defined in IP 71111.13.
b. Findings
No findings were identified.
1R15 Operability Evaluations (71111.15)
a. Inspection Scope
The inspectors reviewed the following issues:
- AR 01202456, NRC Question on RCS [Reactor Coolant System] PIV
[Pressure Isolation Valve] Surveillance Testing;
- AR 01204102, Category A Failure of High Pressure Core Spray [HPCS]
Instrument;
- AR 01219600, Vibration Levels Increased on 0VC04CB;
- AR 01194749, "Division 1 DG Slow Start;" and
- AR 1208215, "1E21F303 Abnormal Flow/Indication During LPCS [Low Pressure
Core Spray] Clearance Hang."
The inspectors selected these potential operability issues based on the risk significance
of the associated components and systems. The inspectors verified that the conditions
did not render the associated equipment inoperable or result in an unrecognized
increase in plant risk. When applicable, the inspectors verified that the licensee
appropriately applied TS limitations, appropriately returned the affected equipment to an
operable status, and reviewed the licensees evaluation of the issue with respect to the
regulatory reporting requirements. 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 evaluation.
In addition, the inspectors verified that problems related to the operability of
safety-related plant equipment were entered into the licensees corrective action
program with the appropriate characterization and significance. Selected action
requests were reviewed to verify that corrective actions were appropriate and
implemented as scheduled.
This inspection constituted five operability evaluation inspection samples as defined in
13 Enclosure
b. Findings
No findings were identified.
1R18 Plant Modifications (71111.18)
a. Inspection Scope
The inspectors reviewed the following temporary plant modification:
- EC 381638, Temporary Modification to Lift Input from A10 Device to A11 Device
for the Division I Diesel Governor.
The inspectors reviewed the temporary modification and the associated 10 CFR 50.59
screening/evaluation against applicable system design basis documents, including the
UFSAR and the TS to verify whether applicable design basis requirements were
satisfied. The inspectors reviewed the Control Room logs and interviewed engineering
and operations department personnel to understand the impact that implementation of
the temporary modification had on operability and availability of the affected plant SSCs.
This inspection constituted one temporary modification inspection sample as defined in
b. Findings
No findings were identified.
1R19 Post-Maintenance Testing (71111.19)
a. Inspection Scope
The inspectors reviewed post-maintenance testing for the following activities to verify
that procedures and test activities were adequate to ensure system operability and
functional capability:
- Work Order 01175527 02, "Replace and Calibrate Capacity Controller
1TCVP013;"
- Work Order 01176252 01, "Test Bus 1A1 Main Feed Breaker Protective Relays;"
- Work Order 01405272-02, 0FP03P Outboard Bearing Elevated Temperature;
- Work Order 01294892-01, Replace the Division 2 Emergency Diesel Generator
LOCA Bypass Relay 3KL4;
- Work Order 01278032-02, Operations PMT [Post-Maintenance Test] for
1E51F030;
- Work Order 01283985-03, Operations PMT for 0VC03CA; and
- Work Order 01298883-01, Reactor Core Isolation Cooling Valve Operability,
(1E51-F079 and F081 only).
The inspectors reviewed the scope of the work performed and evaluated the adequacy
of the specified post-maintenance testing. The inspectors verified that the
14 Enclosure
post-maintenance testing was performed in accordance with approved procedures; that
the procedures contained clear acceptance criteria, which demonstrated operational
readiness and that the acceptance criteria was met; that appropriate test instrumentation
was used; that the equipment was returned to its operational status following testing;
and, that the test documentation was properly evaluated.
In addition, the inspectors verified that problems related to post-maintenance testing
were entered into the licensees corrective action program with the appropriate
characterization and significance. Selected action requests were reviewed to verify that
corrective actions were appropriate and implemented as scheduled.
This inspection constituted seven post-maintenance testing inspection samples as
defined in IP 71111.19.
b. Findings
No findings were identified.
1R22 Surveillance Testing (71111.22)
a. Inspection Scope
The inspectors reviewed the results of the following surveillance testing activities to
determine whether risk-significant systems and equipment were capable of performing
their intended safety function and to verify that the testing was conducted in accordance
with applicable procedural and TS requirements:
- CPS 3822.06, Operation of the Horizontal Fire Pump; (Routine Test)
Air Filter Package Operability Test Run; (Routine Test)
(Inservice Test)
The inspectors observed selected portions of the test activities to verify that the testing
was accomplished in accordance with plant procedures. The inspectors reviewed the
test methodology and documentation to verify that equipment performance was
consistent with safety analysis and design basis assumptions, and that testing
acceptance criteria were satisfied.
In addition, the inspectors verified that surveillance testing problems were entered into
the licensees corrective action program with the appropriate characterization and
significance. Selected action requests were reviewed to verify that corrective actions
were appropriate and implemented as scheduled.
This inspection constituted three in-service tests and two routine surveillance tests for a
total of five inspection samples as defined in IP 71111.22.
15 Enclosure
b. Findings
(1) Failure to Meet Surveillance Testing Requirement for Reactor Coolant System (RCS)
Pressure Isolation Valves (PIVs)
(Closed) Unresolved Item (URI)05000461/2011002-04, Reactor Coolant System
Pressure Isolation Valve Leakage Surveillance Test Procedure Questions
Introduction
The inspectors identified a finding of very low safety significance (Green) with an
associated non-cited violation of TS Surveillance Requirement (TSSR) 3.4.6.1.
The licensee failed to correctly incorporate the required test pressure limits of the TSSR
into the surveillance test procedure and subsequently tested multiple RCS PIVs at
pressures greater than the maximum test pressure of 1025 pounds per square inch
gauge (psig), invalidating the testing.
Discussion
The inspectors reviewed the licensees performance of surveillance testing that was
accomplished in accordance with CPS 9843.01, ISI [Inservice Inspection] Category A
Valve Leak Rate Test, Revision 35. This surveillance test procedure was performed to
satisfy TSSR 3.4.6.1, which required the licensee to verify the equivalent leakage of
each RCS PIV is 0.5 gallon-per-minute (gpm) per nominal inch of valve size up to a
maximum of 5 gpm, at an RCS pressure 1000 psig and 1025 psig in accordance
with the Inservice Testing Program. The licensees Inservice Testing Program specified
testing these valves once every 24-month refueling cycle during an outage.
As described in the Bases for TS 3.4.6.1, the main purpose in establishing a leakage
limit for the RCS PIVs is to prevent overpressure failure of the low pressure portions of
connecting systems. The leakage limit is an indication of whether the PIVs between the
RCS and the connecting systems are degraded or degrading.
During review of CPS 9843.01 and the completed test packages for RCS PIV testing
performed during the last refueling outage, the inspectors noted that much of the testing
was performed at pressures greater than the TSSR 3.4.6.1 maximum test pressure of
1025 psig. The procedure had the test performers calculate a corrected test pressure to
adjust for the elevation differences between the test gage and the valves undergoing
testing. This appeared to be appropriate in order to account for an actual pressure
difference at the valves as read from the test pressure gage to assure that the valves
would be tested at the correct pressure. However, the inspectors found that the test
procedure did not ensure that leakage testing was performed within the 1000-1025 psig
range specified by TSSR 3.4.6.1. Instead of calculating both an upper and a lower test
pressure based on the TSSR 3.4.6.1 limiting pressure range, the procedure had the test
performers calculate only one test pressure based on the maximum limit of 1025 psig.
Step 8.2.4 of the procedure directed the test performers to pressurize the test volume to
1025 psig (+25/-0 psig), rather than 1025 psig (-25/+0 psig). During review of the
completed test packages, the inspectors noted that, not accounting for calculation errors,
test performers pressurized the test volume to the calculated test pressure (+25/-0 psig).
The inspectors noted that the Bases for TS 3.4.6 states that leakage testing at a lower
pressure differential than between the specified maximum RCS pressure and the normal
16 Enclosure
pressure of the connected system during RCS operation (the maximum pressure
differential) is allowed. The observed rate may be adjusted to the maximum pressure
differential by assuming leakage is directly proportional to the pressure differential to the
one-half power. However, the inspectors found that the test procedure did not make any
allowance by way of calculating a corrected leakage for a lower pressure differential.
The inspectors also found no allowance in the TS Bases or in the procedure for testing
with a higher pressure differential.
The inspectors discussed these observations with the licensee and questioned whether
the required test pressure limits of TSSR 3.4.6.1 had been correctly incorporated into
the surveillance test procedure. The inspectors opened URI 05000461/2011002-04
pending additional review and resolution of open questions to determine whether the
surveillance test procedure was adequate to satisfy the surveillance testing requirement.
The licensee initiated action requests AR 01202456 and AR 01212825 to address the
inspectors questions.
In response to the inspectors questions, the licensee discovered that five RCS PIVs
(1E12-F023, 1E12-F042A, 1E12-F042C, 1E21-F006, and 1E22-F005) had been tested
at test pressures greater than the maximum 1025 psig limit specified in TSSR 3.4.6.1.
This resulted in invalid surveillance testing results for these five valves. The inspectors
noted that since the surveillance test procedure was incorrect, it was simply by chance
that only five of the RCS PIVs were found to have been tested above 1025 psig after the
licensee re-calculated corrected test pressures. The licensee performed a risk
assessment of the missed PIV surveillances in accordance with TSSR 3.0.3, which
determined that completion of the surveillances could be delayed up to the 24-month
surveillance interval without a significant increase in plant risk. The inspectors reviewed
the risk assessment and concurred that there was no unacceptable increase in risk.
The licensee also completed an operability evaluation for the TS nonconformance and
concluded that there was reasonable assurance that the affected PIVs were operable
based on engineering judgment. Although there was no defined relationship available to
equate valve seating force to valve seat leakage, the licensee concluded that the
relatively small change (decrease) in seating force due to a relatively small increase in
test pressure above the maximum test pressure would not result in a significant increase
in valve seat leakage such that the limiting leakage rates for the valves would not be
exceeded. The inspectors reviewed the operability evaluation and concluded that the
licensees conclusion was reasonable. The highest corrected test pressure calculated
for a PIV was 7.8 psi higher than the TSSR maximum test pressure.
The inspectors reviewed the licensees apparent cause evaluation for the missed
surveillance. The licensees evaluation highlighted that a missed opportunity to correct
the test pressure discrepancy had occurred in 2005. In December 2004, AR 00282084
was written to identify that the surveillance test procedure would test the RCS PIVs at
pressures up to 25 psig above the maximum pressure specified in TSSR 3.4.6.1.
However, the licensees subsequent evaluation of the described condition completed in
January 2005 was incorrect, in that, it concluded that testing at the higher pressure was
conservative and therefore acceptable. A change was made to CPS 9843.01 as an
enhancement to the procedure to add an explanatory statement accounting for the
apparent test pressure discrepancy. Step 2.1.11 of the test procedure stated, in part,
that [t]o conservatively ensure compliance with TSSR 3.4.6.1 test pressure
requirements, functional differential pressures are established at or above the upper
bound of pressure defined by TSSR 3.4.6.1, recognizing that TSSR 3.4.6.1 Bases state
17 Enclosure
that RCS PIV leakage is directly proportional to pressure to the 1/2 power. However,
according to the Bases for TS 3.4.6 this allowance is only for leakage testing at a lower
pressure differential between the specified maximum RCS pressure and the normal
pressure of the connected system during RCS operation.
The licensee identified that the apparent cause for the incorrect test pressure
specified in the surveillance test procedure was due to a technical human error.
Engineering judgment that testing at a higher pressure was conservative was not
challenged as being outside the literal test pressure band specified in the TSSR.
The inspectors reviewed AR 00282084 and noted that multiple licensee staff had
accepted this flawed engineering judgment, both in Engineering and Operations.
Corrective actions identified by the licensee included changes to CPS 9843.01 to
correct identified discrepancies with the test conditions and acceptance criteria.
In addition, to address a broader issue highlighted by this and other recent inspection
findings involving test control issues, the licensee identified an action from AR 01207467
to evaluate the generic issue involving translation of licensing/design basis requirements
into test procedures. The inspectors considered these corrective actions to be
appropriate.
The licensee stated in the apparent cause evaluation that the technical human error was
made in 2005, prior to the issuance of procedure HU-AA-1212, Technical Task
Risk/Rigor Assessment, Pre-job Brief, Independent Third Party Review, and Post-job
Review, which requires additional measures to be taken to identify assumptions during
engineering work. Therefore, the licensee did not identify any additional corrective
actions to address why the errors were made in the evaluation of AR 00282084.
Analysis
The inspectors determined that the licensees failure to satisfy the surveillance testing
requirement to verify the equivalent leakage of each RCS PIV is 0.5 gpm per nominal
inch of valve size up to a maximum of 5 gpm, at an RCS pressure 1000 psig and
1025 psig was a performance deficiency warranting a significance evaluation.
The inspectors reviewed the examples of minor issues in IMC 0612, Power Reactor
Inspection Reports, Appendix E, Examples of Minor Issues, and found no examples
related to this issue. Consistent with the guidance in IMC 0612, Power Reactor
Inspection Reports, Appendix B, Issue Screening, the inspectors determined that the
finding affected the Initiating Events Cornerstone and was associated with the Procedure
Quality attribute. Specifically, the licensee did not correctly incorporate the required test
pressure limits of TSSR 3.4.6.1 into the surveillance test procedure. This resulted in
testing multiple RCS PIVs at pressures greater than the maximum test pressure of
1025 psig. The inspectors performed a Phase 1 SDP review of this finding using the
guidance provided in IMC 0609, Attachment 0609.04, Phase 1 - Initial Screening and
Characterization of Findings. In accordance with Table 4a, Characterization
Worksheet for IE [Initiating Events], MS [Mitigating Systems], and BI [Barrier Integrity]
Cornerstones, the inspectors determined that that this finding was a licensee
performance deficiency of very low safety significance (Green) because the finding
would not result in exceeding the TS limit for RCS leakage and would not have likely
affected mitigation systems resulting in a loss of safety function. Based on consultation
and review with the Regional Senior Reactor Analyst, the inspectors concluded that the
18 Enclosure
testing deficiency did not result in an increase in valve failure probability or the likelihood
of an initiating event such as an inter-system LOCA.
Cross-Cutting Aspects
The inspectors concluded that because the licensees missed opportunity to correct the
test pressure discrepancy in its surveillance test procedure occurred in January 2005
and no other more recent opportunities reasonably existed to identify and correct the
problem, this issue would not be reflective of current licensee performance and no
cross-cutting aspect was identified.
Enforcement
TSSR 3.4.6.1 requires the licensee to verify the equivalent leakage of each RCS PIV is
0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm, at an RCS
pressure 1000 psig and 1025 psig in accordance with the Inservice Testing Program.
The licensees Inservice Testing Program specified testing these valves once every
24-month refueling cycle during an outage.
Contrary to the above, during surveillance testing between January 12 and 19, 2010,
performed in accordance with CPS 9843.01, ISI Category A Valve Leak Rate Test,
Revision 35, the licensee performed leakage testing of five RCS PIVs (1E12-F023,
1E12-F042A, 1E12-F042C, 1E21-F006, and 1E22-F005) at test pressures greater than
the maximum 1025 psig limit specified in TSSR 3.4.6.1. This resulted in invalid
surveillance testing results for these five valves for the previous 24-month refueling
cycle. Because of the very low safety significance, this violation is being treated as an
non-cited violation consistent with Section 2.3.2 of the NRC Enforcement Policy
(NCV 05000461/2011003-02, Failure to Meet Surveillance Testing Requirement for
Reactor Coolant System Pressure Isolation Valves). The licensee entered this
violation into its corrective action program as AR 01202456 and AR 01212825.
URI 05000461/2011002-04 is closed.
(2) Surveillance Testing of Control Room Ventilation (VC) System
Introduction
The inspectors initiated an Unresolved Item pending additional review and resolution of
open questions to determine whether the licensees VC system monthly operability
surveillance test procedure contained the appropriate requirements and acceptance
limits for intake filtered flow rate from applicable design documents and whether
operators appropriately addressed the operability of VC Train A after identifying a
degraded condition that could have affected the ability of the system to perform its safety
function.
Discussion
The inspectors reviewed the licensees performance of surveillance testing that was
accomplished in accordance with CPS 9070.01, Control Room HVAC Air Filter Package
Operability Test Run, Revision 26d. This surveillance test procedure was performed
to satisfy TSSRs 3.7.3.1 and 3.7.3.2, which required the licensee to operate each
19 Enclosure
VC subsystem with flow through the makeup filter 10 continuous hours with the heater
operating and with flow through the recirculation filter for 15 minutes, respectively.
The surveillance frequency is every 31 days. As described in the Bases for TS 3.7.3,
the ability of the VC system to maintain the habitability of the Control Room envelope is
an explicit assumption for the safety analyses presented in the UFSAR. The high
radiation mode of the VC system is assumed to operate following a design basis
accident. The VC system is designed to maintain a habitable environment in the
Control Room envelope for a 30-day continuous occupancy after a design basis
accident, without exceeding 5 Rem total effective dose equivalent (TEDE) as required
by 10 CFR 50, Appendix A, Criterion 19. The UFSAR Chapter 15 accident analyses
assumed that for a design basis LOCA, the VC system intake filtered flow rate is
3000 +/- 10% cubic feet per minute (cfm).
During testing of VC Train A on April 1, 2011, an operator noted that the filtered make
up flow was oscillating between 2300 and 2880 cfm; however, as stated in Step 8.1.2.h
of the test procedure, flow should have been 2700 to 3300 cfm. The operator annotated
the test procedure with a note stating that the flow was low and initiated AR 01196342 to
have the condition evaluated and corrected. Operators reviewed the acceptance criteria
in Section 9.1 of the test procedure and did not find any upper or lower limits for flow
rate. Operators noted that the Control Room differential pressure remained positive with
the degraded flow condition and, therefore, concluded that VC Train A remained
operable and signed off the completed test procedure as satisfactory with no further
evaluation. Operators did not request a formal operability evaluation from engineering
even though the VC system has a required licensing basis function and the degraded
condition could have affected the ability of the system to perform its safety function.
During review of the completed surveillance test procedure and AR 01196342, the
inspectors questioned: (1) whether VC Train A remained operable with intake filtered
flow less than design, and (2) the absence of an appropriate quantitative acceptance
criterion for filtered flow rate in the test procedure to assure that the system would be
capable of fulfilling its design safety function. The inspectors noted that TSSRs 3.7.3.1
and 3.7.3.2 do not specify upper or lower limits for system intake filtered flow rate, nor
do any other VC system TSSRs. Only the administrative program requirement for
VC system filter testing in TS 5.5.7 specifies a 3000 cfm intake filtered flow rate, but this
testing is performed much less frequently (i.e., every 2 years vice every month).
The inspectors reviewed CPS 9866.01, VG/VC HEPA [High Efficiency Particulate Air]
Filter Leak Test, Revision 26 and noted that this procedure for system filter testing
contained appropriate filtered flow acceptance criteria.
Because the UFSAR Chapter 15 LOCA analyses assumes that the VC system intake
filtered flow rate is 3000 +/- 10%, the inspectors determined that system operability would
be questionable with system flow not within these limits. For determining the radiological
consequences of a design basis LOCA to Control Room operators from external
radiation sources, Calculation C-002, Post LOCA Control Room Operator Dose from
External Sources, Revision 2, assumes the intake filtered flow rate is at the upper limit
of 3300 cfm. The higher value provides a maximum value for iodine buildup in the
charcoal bed under normal conditions. For determining the radiological consequences
of a design basis LOCA using the alternate source term methodology, Calculation
C-020, Reanalysis of Loss of Coolant Accident (LOCA) Using the Alternate Source
Term Methodology, Revision 3, assumes the intake filtered flow rate is 2700 cfm.
Under this analysis, the lower the flow rate the higher the dose to Control Room
20 Enclosure
operators since less filtered air is being provided to the Control Room envelope. Both of
the above calculations support the accident analyses to ensure that post accident dose
to Control Room occupants in the event of a LOCA would be less than 5 Rem TEDE.
The licensee investigated the low flow condition two weeks later on April 15th and
discovered that the VC Train A flow controller was not functioning properly. The flow
controller was replaced with a new one and post-maintenance testing was completed
satisfactorily. The licensee documented the flow controller problem in AR 012003343
and subsequently performed a past operability evaluation. The licensees evaluation
concluded that the system remained operable with the degraded flow condition because
there was sufficient margin in the Control Room post-LOCA dose analysis.
The inspectors reviewed the licensees evaluation and concluded that the results were
reasonable.
In response to the inspectors questions, the licensee initiated AR 01207896 to review
the absence of an appropriate quantitative acceptance criterion for filtered flow rate in
the surveillance test procedure. In addition, the licensee initiated AR 01239007 to
perform an apparent cause evaluation addressing the timeliness of the formal operability
assessment and whether the absence of appropriate acceptance criteria in Section 9.1
of CPS 9070.01 influenced the decision by licensed operators to accept the results of
the surveillance test and not request a formal operability evaluation from engineering
upon discovery of the degraded condition during testing.
At the end of this inspection period, the licensee had just entered this issue into its
corrective action program to investigate the cause and to identify appropriate corrective
actions. This issue is considered to be an Unresolved Item (URI 05000461/2011003-03,
Surveillance Testing of Control Room Ventilation System) pending additional review
and resolution of open questions to determine: (1) whether the surveillance test
procedure contained the appropriate requirements and acceptance limits for VC system
intake filtered flow rate from applicable design documents, and (2) whether operators
appropriately addressed the operability of VC Train A after identifying a degraded
condition that could have affected the ability of the system to perform its safety function.
Cornerstone: Emergency Preparedness
1EP6 Drill Evaluation (71114.06)
.1 Emergency Preparedness Drill Observation
a. Inspection Scope
The inspectors evaluated the conduct of an emergency preparedness drill on May 17,
2011, to identify any weaknesses and deficiencies in classification, notification, and
protective action recommendation development activities. This drill was planned to be
evaluated and was included in performance indicator data regarding drill and exercise
performance. The inspectors observed emergency response operations in the Technical
Support Center to determine whether the event classification, notifications, and
protective action recommendations were performed in accordance with procedures.
The operations simulator was not staffed for this drill. The inspectors also attended the
licensees drill critique to compare any inspector-observed weaknesses with those
identified by the licensees staff in order to evaluate the critique and to verify whether the
21 Enclosure
licensees staff was properly identifying weaknesses and entering them into the
corrective action program.
This inspection constituted one emergency preparedness drill evaluation inspection
sample as defined in IP 71114.06.
b. Findings
No findings were identified.
2. RADIATION SAFETY
Cornerstones: Occupational Radiation Safety
2RS3 In-Plant Airborne Radioactivity Control and Mitigation (71124.03)
This inspection constituted a partial sample as defined in IP 71124.03.
.1 Inspection Planning (02.01)
a. Inspection Scope
The inspectors reviewed the UFSAR to identify areas of the plant designed as potential
airborne radiation areas and any associated ventilation systems or airborne monitoring
instrumentation. Instrumentation review included continuous air monitors (continuous air
monitors and particulate-iodine-noble-gas-type instruments) used to identify changing
airborne radiological conditions such that actions to prevent an overexposure may be
taken. The review included an overview of the respiratory protection program and a
description of the types of devices used. The inspectors reviewed UFSAR, TS, and
emergency planning documents to identify location and quantity of respiratory protection
devices stored for emergency use.
Inspectors reviewed the licensees procedures for maintenance, inspection, and use of
respiratory protection equipment including self-contained breathing apparatus (SCBA) as
well as procedures for air quality maintenance.
The inspectors reviewed reported performance indicators to identify any related to
unintended dose resulting from intakes of radioactive material.
b. Findings
Introduction
The inspectors identified a discrepancy between the SCBA configuration and the
Operating and Instruction Manual. Specifically, the licensee procedure for maintaining
the respiratory equipment did not specify the authorized battery and the licensee used
batteries other than those specified in the Operating and Instruction Manual.
22 Enclosure
Discussion
The licensee used MSA MMR Air Mask with Firehawk Regulator SCBA units.
This model is National Institute for Occupational Safety and Health (NIOSH) approved
and includes a heads up display to inform the user of the amount of air remaining in the
tank through a series of light emitting diodes and is powered by a series of batteries.
The Operating and Instruction Manual includes the NIOSH approval for the equipment
as well as the cautions and limitations for that approval. Item N states that
Never substitute, modify, add, or omit parts. Use only exact replacement parts in the
configuration as specified by the manufacturer. The manufacturer includes a caution to
[U]se only Duracell NEDA 24A or Eveready NEDA 24AC AAA alkaline batteries. Use of
other batteries will void the Intrinsic Safety approval. Additionally, the heads up display
units were labeled with a similar message; however, the batteries listed were different.
Consequently there was another discrepancy between the SCBA Operating and
Instruction Manual and the manufacturer labeling on the equipment. The inspectors
identified that Rayovac batteries, not listed in the Operating and Instruction Manual or
the labels, were used in the SCBA units. The licensee was attempting to obtain
clarification from the manufacturer for the correct batteries and impact of using other
batteries. The issue remains under review by the NRC and is categorized as an
Unresolved Item pending completion of that revised evaluation and NRC review
(URI 05000461/2011003-04, NIOSH Approval of SCBAs).
.2 Engineering Controls (02.02)
a. Inspection Scope
The inspectors assessed whether the licensee had established trigger points
(e.g., the Electric Power Research Institutes Alpha Monitoring Guidelines for
Operating Nuclear Power Stations) for evaluating levels of airborne beta-emitting
(e.g., plutonium-241) and alpha-emitting radionuclides.
b. Findings
No findings were identified.
.3 Use of Respiratory Protection Devices (02.03)
a. Inspection Scope
The inspectors assessed whether respiratory protection devices used to limit the intake
of radioactive materials were certified by the National Institute for Occupational Safety
and Health/Mine Safety and Health Administration or have been approved by the
NRC per 10 CFR 20.1703(b).
The inspectors reviewed records of air testing for supplied-air devices and self-contained
breathing apparatus bottles to assess whether the air used in these devices meets or
exceeds Grade D quality. The inspectors reviewed plant breathing air supply systems to
determine whether they meet the minimum pressure and airflow requirements for the
devices in use.
23 Enclosure
The inspectors selected several individuals qualified to use respiratory protection
devices, and assessed whether they have been deemed fit to use the devices by a
physician.
The inspectors selected several individuals assigned to wear a respiratory protection
device and observed them donning, doffing, and functionally checking the device as
appropriate.
The inspectors chose multiple respiratory protection devices staged and ready for use
in the plant or stocked for issuance for use. The inspectors assessed the physical
condition of the device components (mask or hood, harnesses, air lines, regulators,
air bottles, etc.) and reviewed records of routine inspection for each. The inspectors
selected several of the devices and reviewed records of maintenance on the vital
components (e.g., pressure regulators, inhalation/exhalation valves, hose couplings).
b. Findings
No findings were identified.
.4 Self-Contained Breathing Apparatus for Emergency Use (02.04)
a. Inspection Scope
Based on the UFSAR, TS, and emergency operating procedure requirements, the
inspectors reviewed the status and surveillance records of SCBAs staged in-plant for
use during emergencies. The inspectors reviewed the licensees capability for refilling
and transporting SCBA air bottles to and from the Control Room and Operations Support
Center during emergency conditions.
The inspectors selected several individuals on Control Room shift crews and from
designated departments currently assigned emergency duties (e.g., onsite search and
rescue duties) to assess whether Control Room operators and other emergency
response and radiation protection personnel (assigned in-plant search and rescue duties
or as required by emergency operating procedures or the Emergency Plan) were trained
and qualified in the use of SCBAs (including personal bottle change out). The inspectors
evaluated whether personnel assigned to refill bottles were trained and qualified for that
task.
The inspectors determined whether appropriate mask sizes and types are available for
use (i.e., in-field mask size and type match what was used in fit-testing). The inspectors
determined whether on-shift operators had no facial hair that would interfere with the
sealing of the mask to the face and whether vision correction (e.g., glasses inserts or
corrected lenses) was available as appropriate.
The inspectors reviewed the past two years of maintenance records for select SCBA
units used to support operator activities during accident conditions and designated as
ready for service to assess whether any maintenance or repairs on any SCBA units
vital components were performed by an individual, or individuals, certified by the
manufacturer of the device to perform the work. The vital components typically are the
pressure-demand air regulator and the low-pressure alarm. The inspectors reviewed the
onsite maintenance procedures governing vital component work to determine any
24 Enclosure
inconsistencies with the SCBA manufacturers recommended practices. For those
SCBAs designated as ready for service, the inspectors determined whether the
required, periodic air cylinder hydrostatic testing was documented and up to date, and
the retest air cylinder markings required by the U.S. Department of Transportation were
in place.
b. Findings
Introduction
The inspectors identified missing spectacle kits for one licensed operator that was
required to wear corrective lenses while performing licensed activities. Licensee
procedure RP-AA-440, Respiratory Protection Program states, An individual who
requires vision correction and may need to wear full facepiece-type respirators is
required to obtain the appropriate lenses/spectacle kit, unless the individual is able to
wear contact lenses.
Discussion
Spectacle kits are corrective lenses designed to fit inside a respirator that allow the user
to wear corrective lenses without compromising the seal integrity of the respirator.
A user that requires corrective lenses to complete work activities and does not have a
spectacle kit could not perform the work while wearing the respirator. The licensee
maintains a central location for licensed operators to store spectacle kits and a separate
storage location for members of the fire brigade. These storage locations ensure that
the spectacle kits are centrally located to facilitate a rapid response when required
(fires, chemical spills, or other emergencies).
The licensee did not have a validation process to ensure that individuals who may need
to wear full face-piece-type respirators actually have the required spectacle kits.
The licensee indicated that a review/evaluation of this issue would be completed.
The issue remains under review by the NRC and is categorized as an Unresolved Item
pending completion of that revised evaluation and NRC review
(URI 05000461/2011003-05, Missing Respirator Spectacle Kits).
.5 Problem Identification and Resolution (02.05)
a. Inspection Scope
The inspectors evaluated whether problems associated with the control and mitigation of
in-plant airborne radioactivity were being identified by the licensee at an appropriate
threshold and were properly addressed for resolution in the licensee corrective action
program. The inspectors assessed whether the corrective actions were appropriate for a
selected sample of problems involving airborne radioactivity and were appropriately
documented by the licensee.
b. Findings
No findings were identified.
25 Enclosure
2RS4 Occupational Dose Assessment (71124.04)
This inspection constituted a partial sample as defined in IP 71124.04.
.1 Inspection Planning (02.01)
a. Inspection Scope
The inspectors reviewed the results of radiation protection program audits related to
internal and external dosimetry (e.g., licensees quality assurance audits,
self-assessments, or other independent audits) to gain insights into overall licensee
performance in the area of dose assessment and focus the inspection activities
consistent with the principle of smart sampling.
The inspectors reviewed the most recent National Voluntary Laboratory Accreditation
Program accreditation report on the vendors most recent results to determine the status
of the contractors accreditation.
A review was conducted of the licensee procedures associated with dosimetry
operations, including issuance/use of external dosimetry (routine, multi-badging,
extremity, neutron, etc.), assessment of internal dose (operation of whole body counter,
assignment of dose based on derived air concentration-hours, urinalysis, etc.), and
evaluation of and dose assessment for radiological incidents (distributed contamination,
hot particles, loss of dosimetry, etc.).
The inspectors evaluated whether the licensee had established procedural requirements
for determining when external and internal dosimetry is required.
b. Findings
No findings were identified.
.2 External Dosimetry (02.02)
a. Inspection Scope
The inspectors evaluated the onsite storage of dosimeters before their issuance, during
use, and before processing/reading. The inspectors also reviewed the guidance
provided to radiation workers with respect to care and storage of dosimeters.
The inspectors assessed whether non-National Voluntary Laboratory Accreditation
Program accredited passive dosimeters (e.g., direct ion storage sight read dosimeters)
were used according to licensee procedures that provide for periodic calibration,
application of calibration factors, usage, reading (dose assessment) and zeroing.
The inspectors assessed the use of active dosimeters (electronic personal dosimeters)
to determine if the licensee uses a correction factor to address the response of the
electronic personal dosimeter as compared to the passive dosimeter for situations when
the electronic personal dosimeter must be used to assign dose and whether the
correction factor is based on sound technical principles.
26 Enclosure
The inspectors reviewed dosimetry occurrence reports or corrective action program
documents for adverse trends related to electronic personal dosimeters, such as
interference from electromagnetic frequency, dropping or bumping, failure to hear
alarms, etc. The inspectors assessed whether the licensee had identified any trends
and implemented appropriate corrective actions.
b. Findings
No findings were identified.
.3 Internal Dosimetry (02.03)
Routine Bioassay (In Vivo)
a. Inspection Scope
The inspectors reviewed procedures used to assess the dose from internally deposited
nuclides using whole body counting equipment. The inspectors evaluated whether the
procedures addressed methods for differentiating between internal and external
contamination, the release of contaminated individuals, the route of intake, and the
assignment of dose.
The inspectors reviewed the whole body count process to determine if the frequency of
measurements was consistent with the biological half-life of the nuclides available for
intake.
The inspectors reviewed the licensee's evaluation for use of its portal radiation monitors
as a passive monitoring system to determine if instrument minimum detectable activities
were adequate to determine the potential for internally deposited radionuclides sufficient
to prompt additional investigation.
The inspectors selected several whole body counts and evaluated whether the counting
system used had sufficient counting time/low background to ensure appropriate
sensitivity for the potential radionuclides of interest. The inspectors reviewed the
radionuclide library used for the count system to determine its appropriateness.
The inspectors evaluated whether any anomalous count peaks/nuclides indicated in
each output spectra received appropriate disposition. The inspectors reviewed the
licensee's 10 CFR 61 data analyses to determine whether the nuclide libraries included
appropriate gamma-emitting nuclides. The inspectors evaluated how the licensee
accounts for hard-to-detect nuclides in the dose assessment.
b. Findings
No findings were identified.
Special Bioassay (In Vitro)
a. Inspection Scope
There were no internal dose assessments obtained using in vitro monitoring for the
inspectors to review. The inspectors reviewed and assessed the adequacy of the
licensees program for in vitro monitoring (i.e., urinalysis and fecal analysis) of
27 Enclosure
radionuclides (tritium, fission products, and activation products), including collection and
storage of samples.
The inspectors reviewed the vendor laboratory quality assurance program and assessed
whether the laboratory participated in an industry recognized cross-check program
including whether out-of-tolerance results were resolved appropriately.
b. Findings
No findings were identified.
Internal Dose Assessment - Airborne Monitoring
a. Inspection Scope
The inspectors reviewed the licensee's program for airborne radioactivity assessment
and dose assessment, as applicable, based on airborne monitoring and calculations of
derived air concentration. The inspectors determined whether flow rates and collection
times for air sampling equipment were adequate to allow lower limits of detection to be
obtained. The inspectors also reviewed the adequacy of procedural guidance to assess
internal dose if respiratory protection was used. The licensee had not performed dose
assessments using airborne/derived air concentration monitoring since the last
inspection.
b. Findings
No findings were identified.
Internal Dose Assessment - Whole Body Count Analyses
a. Inspection Scope
The inspectors reviewed several dose assessments performed by the licensee using the
results of whole body count analyses. The inspectors determined whether affected
personnel were properly monitored with calibrated equipment and that internal
exposures were assessed consistent with the licensee's procedures.
b. Findings
No findings were identified.
.4 Special Dosimetric Situations (02.04)
Declared Pregnant Workers
a. Inspection Scope
The inspectors assessed whether the licensee informed workers, as appropriate, of the
risks of radiation exposure to the embryo/fetus, the regulatory aspects of declaring a
pregnancy, and the specific process to be used for (voluntarily) declaring a pregnancy.
28 Enclosure
The inspectors selected individuals who had declared pregnancy during the current
assessment period and evaluated whether the licensees radiological monitoring
program (internal and external) for declared pregnant workers was technically adequate
to assess the dose to the embryo/fetus. The inspectors reviewed exposure results and
monitoring controls employed by the licensee with respect to the requirements of
b. Findings
No findings were identified.
Dosimeter Placement and Assessment of Effective Dose Equivalent for External
Exposures
a. Inspection Scope
The inspectors reviewed the licensee's methodology for monitoring external dose in
non-uniform radiation fields or where large dose gradients exist. The inspectors
evaluated the licensee's criteria for determining when alternate monitoring, such as use
of multi-badging, was to be implemented.
The inspectors reviewed dose assessments performed using multi-badging to evaluate
whether the assessment was performed consistently with licensee procedures and
dosimetric standards.
b. Findings
No findings were identified.
Shallow Dose Equivalent
a. Inspection Scope
The inspectors reviewed shallow dose equivalent dose assessments for adequacy.
The inspectors evaluated the licensees method (e.g., VARSKIN or similar code) for
calculating shallow dose equivalent from distributed skin contamination or discrete
radioactive particles.
b. Findings
No findings were identified.
Neutron Dose Assessment
a. Inspection Scope
The inspectors evaluated the licensees neutron dosimetry program, including dosimeter
types and/or survey instrumentation.
The inspectors reviewed neutron exposure situations (e.g., independent spent fuel
storage installation operations or at-power containment entries) and assessed whether:
(a) dosimetry and/or instrumentation was appropriate for the expected neutron spectra;
29 Enclosure
(b) there was sufficient sensitivity for low dose and/or dose rate measurement; and
(c) neutron dosimetry was properly calibrated. The inspectors also assessed whether
interference by gamma radiation had been accounted for in the calibration and whether
time and motion evaluations were representative of actual neutron exposure events, as
applicable.
b. Findings
No findings were identified.
Assigning Dose of Record
a. Inspection Scope
For the special dosimetric situations reviewed in this section, the inspectors assessed
how the licensee assigns dose of record for total effective dose equivalent, shallow dose
equivalent, and lens dose equivalent. This included an assessment of external and
internal monitoring results, supplementary information on Individual exposures
(e.g., radiation incident investigation reports and skin contamination reports), and
radiation surveys and/or air monitoring results when dosimetry was based on these
techniques.
b. Findings
No findings were identified.
.5 Problem Identification and Resolution (02.05)
a. Inspection Scope
The inspectors assessed whether problems associated with occupational dose
assessment were being identified by the licensee at an appropriate threshold and
were properly addressed for resolution in the licensees corrective action program.
The inspectors assessed the appropriateness of the corrective actions for a selected
sample of problems documented by the licensee involving occupational dose
assessment.
b. Findings
No findings were identified.
4. OTHER ACTIVITIES
4OA1 Performance Indicator Verification (71151)
.1 Review of Submitted Quarterly Data
a. Inspection Scope
The inspectors performed a review of the data submitted by the licensee for the First
Quarter 2011 Performance Indicators for any obvious inconsistencies prior to its public
release in accordance with IMC 0608, "Performance Indicator Program."
30 Enclosure
This inspection was not considered to be an inspection sample as defined in IP 71151.
b. Findings
No findings were identified.
.2 Mitigating Systems Performance Index - Emergency Alternating Current (AC)
Power System
a. Inspection Scope
The inspectors reviewed a sample of plant records and data against the reported
Mitigating Systems Performance Index (MSPI) - Emergency AC Power System
Performance Indicator. To determine the accuracy of the performance indicator data
reported, performance indicator definitions and guidance contained in Nuclear Energy
Institute (NEI) 99-02, Regulatory Assessment Performance Indicator Guideline,
Revision 6, were used. The inspectors reviewed the MSPI derivation reports,
Control Room logs, Maintenance Rule database, Licensee Event Reports (LERs), and
maintenance and test data from July 2010 through March 2011, to validate the accuracy
of the performance indicator data reported. The inspectors reviewed the MSPI
component risk coefficient to determine if it had changed by more than 25% in value
since the previous inspection, and if so, that the change was in accordance with
applicable NEI guidance. The inspectors also reviewed the licensees corrective action
program database to determine if any problems had been identified with the
performance indicator data collected or transmitted for this performance indicator.
This inspection constituted one MSPI - Emergency AC Power System Performance
Indicator verification inspection sample as defined in IP 71151.
b. Findings
No findings were identified.
4OA2 Identification and Resolution of Problems (71152)
.1 Routine Review of Identification and Resolution of Problems
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 that they were
being entered into the licensees corrective action program at an appropriate threshold,
that adequate attention was being given to timely corrective actions, and that adverse
trends were identified and addressed. Some minor issues were entered into the
licensees corrective action program as a result of the inspectors observations; however,
they are not discussed in this report.
This inspection was not considered to be an inspection sample as defined in IP 71152.
31 Enclosure
b. Findings
No findings were identified.
.2 Annual In-Depth Review Sample
a. Inspection Scope
The inspectors selected the following action request for in-depth review:
- AR 01017464, 1B21F028A: 9861.04 LLRT [Local Leak Rate Test] on MSL
[Main Steam Line] A, B, and C Test Failure.
The inspectors verified the following attributes during their review of the licensee's
corrective actions for the above action requests and other related action requests:
- Complete and accurate identification of the problem in a timely manner
commensurate with its safety significance and ease of discovery;
- Consideration of the extent of condition, generic implications, common cause and
previous occurrences;
- Evaluation and disposition of operability/reportability issues;
- Classification and prioritization of the resolution of the problem, commensurate
with safety significance;
- Identification of the root and contributing causes of the problem; and
- Identification of corrective actions, which were appropriately focused to correct
the problem.
The inspectors discussed the corrective actions and associated action request
evaluation with licensee personnel.
This inspection constituted one annual in-depth review sample as defined in IP 71152.
b. Findings and Observations
No findings were identified.
4OA3 Followup of Events and Notices of Enforcement Discretion (71153)
.1 (Closed) LER 05000461/2009-005-01, Manual Scram on High Water Level Due to
Reactor Recirc [Recirculation] Pump Trip, Supplement 1
On October 15, 2009, Unit 1 was manually scrammed following an unexpected trip
of the B reactor recirculation pump. Operators manually scrammed the reactor just
before reactor vessel water level reached the Level 8 (high level) reactor scram set
point. After the unit was shut down, the licensee identified that the pump motor had
failed due to an internal electrical fault. The licensee reported this event in
LER 05000461/2009-005-00 as a condition that resulted in the manual actuation of the
reactor protection system in accordance with 10 CFR 50.73(a)(2)(iv)(A).
The performance issue related to this event was discussed in NRC Inspection Report
05000461/2010-002. The inspectors documented a finding of very low safety
significance as a result of the licensees failure to correct a non-conforming condition
32 Enclosure
with inadequate feedwater level control system response that resulted in a second
reactor scram for the same cause.
The licensee submitted Supplement 1 to the original LER to revise the cause for the
reactor vessel water level control issue and to update corrective actions that were
completed. The inspectors determined that the information provided in
LER 05000461/2009-005-01 did not change the conclusion of the previous review.
LER 05000461/2009-005-01 is closed.
This inspection constituted one event followup inspection sample as defined in IP 71153.
.2 (Closed) LER 05000461/2008-001-02, Reactor Recirc [Recirculation] Pump Trip
Initiates Automatic Scram on High RPV [Reactor Pressure Vessel] Water Level,
Supplement 2
On February 10, 2008, Unit 1 automatically scrammed following an unexpected trip of
the B reactor recirculation pump when reactor vessel water level reached the Level 8
(high level) reactor scram set point. The licensee reported this event in
LER 05000461/2008-001-00 and LER 05000461/2008-001-01 as a condition that
resulted in the automatic actuation of the reactor protection system in accordance with
10 CFR 50.73(a)(2)(iv)(A). The performance issues related to this event were discussed
in NRC Inspection Report 05000461/2008-004. The inspectors concluded that the
licensees failure to perform adequate post-maintenance testing (i.e., feedwater level
control system tuning) following the replacement of a feedwater level control system
dynamic compensator card during the Cycle 10 refueling outage in February 2006 was a
finding of very low safety significance. The inspectors also concluded that the licensees
failure to evaluate an unexpected and unknown cause for stray voltage in the
End-of-Cycle Recirculation Pump Trip circuit discovered during post-modification testing
during the Cycle 11 refueling outage in February 2008 was a finding of very low safety
significance.
The licensee submitted Supplement 2 to the original LER to revise the cause for the
reactor vessel water level control issue and to update corrective actions that were
completed. The inspectors determined that the information provided in
LER 05000461/2008-001-02 did not change the conclusion of the previous review.
LER 05000461/2008-001-02 is closed.
This inspection constituted one event followup inspection sample as defined in IP 71153.
4OA5 Other Activities
.1 (Closed) NRC Temporary Instruction (TI) 2515/183, Followup to the Fukushima Daiichi
Nuclear Station Fuel Damage Event
The inspectors assessed the activities and actions taken by the licensee to assess its
readiness to respond to an event similar to the Fukushima Daiichi Nuclear Plant fuel
damage event. This included: (1) an assessment of the licensees capability to mitigate
conditions that may result from beyond design basis events, with a particular emphasis
on strategies related to the spent fuel pool, as required by NRC Security Order
Section B.5.b issued on February 25, 2001, as committed to in Severe Accident
Management Guidelines, and as required by 10 CFR 50.54(hh); (2) an assessment of
33 Enclosure
the licensees capability to mitigate station blackout conditions, as required by
10 CFR 50.63 and station design bases; (3) an assessment of the licensees capability
to mitigate internal and external flooding events, as required by station design bases;
and (4) an assessment of the thoroughness of the walkdowns and inspections of
important equipment needed to mitigate fire and flooding events, which were performed
by the licensee to identify any potential loss of function of this equipment during seismic
events possible for the site.
Inspection Report 05000461/2011011 (ML111320336) documented detailed results of
this inspection activity.
.2 (Closed) NRC TI 2515/184, Availability and Readiness Inspection of Severe Accident
Management Guidelines (SAMGs)
On May 20, 2011, the inspectors completed a review of the licensees Severe Accident
Management Guidelines (SAMGs), implemented as a voluntary industry initiative in the
1990s, to determine (1) whether the SAMGs, were available and updated, (2) whether
the licensee had procedures and processes in place to control and update its SAMGs,
(3) the nature and extent of the licensees training of personnel on the use of SAMGs,
and (4) the licensee personnels familiarity with SAMG implementation.
The results of this review were provided to the NRC task force chartered by the
Executive Director for Operations to conduct a near-term evaluation of the need for
agency actions following the Fukushima Daiichi fuel damage event in Japan.
Plant-specific results for Clinton Power Station were provided as an Enclosure to a
memorandum to the Chief, Reactor Inspection Branch, Division of Inspection and
Regional Support, dated June 1, 2011, (ML111520396).
.3 (Closed) NRC TI 2515/177, Managing Gas Accumulation in Emergency Core Cooling,
Decay Heat Removal, and Containment Spray Systems (NRC Generic Letter 2008-01)
a. Inspection Scope
The inspectors verified that the onsite documentation, system hardware, and licensee
actions were consistent with the information provided in the licensees response to
NRC Generic Letter (GL) 2008-01, Managing Gas Accumulation in Emergency Core
Cooling (ECCS), Decay Heat Removal (DHR), and Containment Spray Systems.
Specifically, the inspectors verified that the licensee has implemented or was in the
process of implementing the commitments, modifications, and programmatically
controlled actions described in the licensees response to GL 2008-01. The inspection
was conducted in accordance with TI 2515/177, Managing Gas Accumulation in
Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems
(NRC Generic Letter 2008-01), and considered the site-specific supplemental
information provided by the Office of Nuclear Reactor Regulation (NRR) to the
inspectors.
The documents reviewed are listed in the Attachment to this report.
34 Enclosure
b. Inspection Documentation
The selected TI areas of inspection were licensing basis, design, testing, and corrective
actions. The documentation of the inspection effort and any resulting observations are
below.
Licensing Basis
The inspectors reviewed selected portions of licensing basis documents to verify that
they were consistent with the NRR assessment report and that they were processed by
the licensee. The licensing basis verification included the verification of selected
portions of TS, TS Bases, UFSAR, and Operations Requirements Manual. The
inspectors also verified that applicable documents that described the plant and plant
operation, such as calculations, piping and instrumentation diagrams (P&IDs),
procedures, and corrective action program documents, addressed the areas of concern
and were changed if needed following plant changes. The inspectors also confirmed
that the frequency of selected surveillance procedures were at least as frequent as
required by TSs. Finally, the inspectors confirmed that: (1) the licensee will review and
evaluate the resolution of TS issues with respect to the changes contained in the
Technical Specification Task Force (TSTF) traveler following NRC approval; and (2) that
a license amendment request will be submitted to the NRC within 180 days following the
evaluation, if necessary. The completion date for this regulatory commitment is
contingent upon the approval of the TSTF.
Design
The inspectors reviewed selected design documents, performed system walkdowns, and
interviewed plant personnel to verify that the design and operating characteristics were
addressed by the licensee. Specifically:
- The inspectors assessed the licensees void acceptance criteria and noted that the
licensee established void volume acceptance criteria for piping locations located at
system high points to be used during field verifications. The void volumes were
derived based on pipe internal diameter and as-built slope, and internal height of the
void.
- The inspectors selectively reviewed applicable documents, including calculations,
engineering evaluations, and vendor technical manuals, with respect to gas
accumulation in the residual heat removal (RHR) and high pressure core spray
(HPCS) systems. Specifically, the inspectors verified that these documents
addressed venting requirements, keep-full systems, and void control during system
realignments.
- The inspectors conducted a walkdown of selected accessible portions of the RHR
system in sufficient detail to assess the licensees walkdowns. The inspectors also
verified that the information obtained during the licensees walkdown was consistent
with the items identified during the inspectors independent walkdown. In addition,
the inspectors verified that the licensee had P&IDs and isometric drawings that
describe the RHR system configurations and had confirmed the accuracy of the
drawings. The inspectors reviewed selected portions of isometric drawings and
considered the following:
35 Enclosure
a. High point vents were identified.
b. High points that do not have vents were recognizable.
c. Other areas where gas can accumulate and potentially impact subject system
operability, such as at orifices in horizontal pipes, isolated branch lines, heat
exchangers, improperly slopped piping, and under closed valves, were described
in the drawings or in referenced documentation.
d. Horizontal pipe centerline elevation deviations and pipe slopes in nominally
horizontal lines that exceed specified criteria were identified.
e. All pipes and fittings were clearly shown.
f. The drawings were up-to-date with respect to recent hardware changes and that
any discrepancies between as-built configurations and the drawings were
documented and entered into the corrective action program for resolution.
- The inspectors also conducted similar walkdowns of selected inaccessible portions
of the HPCS system in other inspection periods. These additional activities counted
toward the completion of this TI and were documented in NRC Inspection Report
- The inspectors verified that licensee walkdowns have been completed. In addition,
the inspectors selectively verified that information obtained during the licensees
walkdowns was addressed and incorporated into procedures, the corrective actions
program, and the void management program.
Testing
The inspectors reviewed selected surveillance, post-modification test, and
post-maintenance test procedures and results to verify that the licensee has approved
and was using procedures that were adequate to address the issue of gas accumulation
and/or intrusion in the subject systems. This review included the verification of
procedures used for conducting surveillances and determination of void volumes to
ensure that the void criteria was satisfied and will be reasonably ensured to be satisfied
until the next scheduled void surveillance. Also, the inspectors reviewed procedures
used for filling and venting following conditions which may have introduced voids into the
subject systems to verify that the procedures addressed testing for such voids and
provided processes for their reduction or elimination. Additionally, the inspectors
reviewed ultrasonic test results for locations that were determined to be susceptible to
voiding based on the licensees walkdowns and evaluations.
The inspectors also review selected portions of procedures used during the surveillance
testing of subject systems in a separate inspection activity. This additional activity
counted towards the completion of this TI and was documented in NRC Inspection
Report 05000461/2010004.
Corrective Actions
The inspectors reviewed selected licensee assessment reports and corrective action
program documents to assess the effectiveness of the licensees corrective actions
36 Enclosure
when addressing the issues associated with GL 2008-01. The inspectors identified one
instance where the licensee failed to implement follow-up void management activities
when an action tracking item to resolve the issue was cancelled. Specifically, the
licensee cancelled a modification that was initiated to address a void identified in RHR
system piping with a horizontal centerline elevation deviation. When the modification
was cancelled, the licensee did not establish a follow-up void management activity.
The licensee performed a confirmatory ultrasonic examination and determined no void
existed. The licensee also issued AR 01212387 to document the inspectors concern of
canceling an action without evaluating the need for a substitute action.
In addition, the inspectors verified that selected corrective actions identified in the
licensees nine-month and supplemental reports were documented. The inspectors also
conducted a similar review of corrective action program documents in a separate
inspection activity. This additional activity counted towards the completion of this TI and
was documented in NRC Inspection Report 05000461/2010-004.
c. Findings
No findings were identified. Based on this review, the inspectors concluded that there is
reasonable assurance that the licensee will complete all outstanding items and
incorporate this information into the design basis and operational practices. Therefore,
this TI is considered closed.
4OA6 Management Meetings
.1 Resident Inspectors Exit Meeting
The inspectors presented the inspection results to Mr. W. Noll and other members of the
licensees staff at the conclusion of the inspection on July 13, 2011. The licensee
acknowledged the findings presented. Proprietary information was examined during this
inspection, but is not specifically discussed in this report.
.2 Interim Exit Meetings
Interim exit meetings were conducted for:
- The results of NRC TI 2515/184, Availability and Readiness Inspection of Severe
Accident Management Guidelines (SAMGs), inspection with Mr. F. Kearney and
other members of the licensees staff at the conclusion of the inspection on
May 20, 2011. The inspector confirmed that none of the potential report input
discussed was considered proprietary.
- The results of the Occupational Dose Assessment and In-Plant Airborne
Radioactivity Control and Mitigation inspection with Mr. F. Kearney and other
members of the licensees staff at the conclusion of the inspection on May 13, 2011,
and subsequently with Mr. J. Stovall on June 16, 2011. The inspector confirmed that
none of the potential report input discussed was considered proprietary.
- The results of the Triennial Heat Sink and NRC TI 2515/177, Managing Gas
Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment
Spray Systems (NRC Generic Letter 2008-01), inspection with Mr. F. Kearney and
37 Enclosure
other members of the licensee staff on May 6, 2011, and subsequently with
Mr. R. Frantz on July 15, 2011. The inspectors confirmed that none of the potential
report input discussed was considered proprietary.
.3 Regulatory Performance Meeting
On April 28, 2011, the NRC held a meeting with the licensee at the Clinton Power
Station to discuss the Clinton Power Station annual plant performance assessment.
The assessment results were previously documented in Inspection Report
.4 Public Meeting
On April 28, 2011, the NRC held a public open house meeting at the Clinton Elks Lodge
to engage interested members of the public on the performance of the Clinton Power
Station and the role of the NRC in ensuring safe plant operations upon completion of the
Clinton Power Station annual plant performance assessment in accordance with
Section 09.01 of IMC 0305, Operating Reactor Assessment Program.
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 meets the criteria of Section 2.3.2 of the
NRC Enforcement Policy, NUREG-1600, for being dispositioned as an non-cited
violation.
.1 Failure to Evaluate the Effects of Dynamic Loads on the Containment Spray Piping
Based on an issue identified at another facility, the licensee initiated AR 01197314 to
verify that the normally voided section of the RHR system containment spray piping
had been properly analyzed for dynamic loading during spray initiation. The licensee
determined that General Electric Design Specification 22A3139, Residual Heat
Removal, specified a dynamic loading analysis, which was required by American
Society of Mechanical Engineers Code,Section III, for the normally voided section of
piping; however, the licensee was unable to locate the analyses or justification for not
performing the analyses. The licensees initial evaluation as documented in calculation
3C10-0175-001, Design and Analysis of Clinton Containment Spray System, was
based on a review of Electric Power Research Institute Topical Report (TR)-106438,
Water Hammer Handbook for Nuclear Plant Engineers and Operators. The evaluation
concluded that a severe water hammer would not occur based on the condition that the
valves required to open were slow acting. However, the licensee determined the
evaluation was based on the case of a valve closing and not on the actual condition of
a valve opening and a slug of water moving through the voided section of pipe.
The licensee then performed an analysis of a water slug and confirmed the system
remained operable. Failure to have a dynamic loading analysis for this piping as
required by the design specification was considered a violation of 10 CFR 50,
Appendix B, Criterion III, Design Control. This violation was not greater than Green
because the dynamic loading was verified to be within the capability of the piping design.
ATTACHMENT: SUPPLEMENTAL INFORMATION
38 Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
K. Baker, Design Engineering Senior Manager
R. Campbell, RP Technical Specialist
T. Chalmers, Operations Director
C. Culp, Engineering
J. Cunningham, Security Manager
B. Davis, Regulatory Assurance Manager
J. Domitrovich, Work Management Director
C. Dunn, Shift Operations Superintendent
S. Fatora, Maintenance Director
R. Frantz, Regulatory Assurance
S. Gackstetter, Training Director
M. Heger, Mechanical/Structural Design Engineering Manager
N. Hightower, Radiological Engineering Manager
F. Kearney, Site Vice President
D. Kemper; Plant Engineering Senior Manager
A. Khanifar, Engineering Director
M. Kimmich, Engineering
S. Lakebrink, Mechanical Design Engineering
K. Leffel, Operations Support Manager
W. Noll, Site Vice President
J. Peterson, Regulatory Assurance
C. Rocha, Nuclear Oversight Manager
S. Soliman, Senior Chemist
J. Stovall, Radiation Protection Manager
B. Taber, Plant Manager
J. Ufert, Fire Marshall
T. Veitch, Chemistry Manager
1 Attachment
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
Opened
05000461/2011-003-01 NCV Deficiencies with RCIC Room Heat Up Analyses
05000461/2011-003-02 NCV Failure to Meet Surveillance Testing Requirement for
Reactor Coolant System Pressure Isolation Valves
(Section 1R22.b.(1))
05000461/2011-003-03 URI Surveillance Testing of Control Room Ventilation System
(Section 1R22.b.(2))
05000461/2011-003-04 URI NIOSH Approval of SCBAs (Section 2RS3.1)
05000461/2011-003-05 URI Missing Respirator Spectacle Kits (Section 2RS3.4)
Closed
05000461/2011-003-01 NCV Deficiencies with RCIC Room Heat Up Analyses
05000461/2011-003-02 NCV Failure to Meet Surveillance Testing Requirement for
Reactor Coolant System Pressure Isolation Valves
(Section 1R22.b.(1))
05000461/2011-002-04 URI Reactor Coolant System Pressure Isolation Valve Leakage
Surveillance Test Procedure Questions (Section 1R22.b.(1))
05000461/2009-005-01 LER Manual Scram on High Water Level Due to Reactor Recirc
[Recirculation] Pump Trip, Supplement 1 (Section 4OA3.1)
05000461/2008-001-02 LER Reactor Recirc [Recirculation] Pump Trip Initiates Automatic
Scram on High RPV [Reactor Pressure Vessel] Water Level
(Section 4OA3.2)
2515/183 TI Followup to the Fukushima Daiichi Nuclear Station Fuel
Damage Event (Section 4OA5.1)
2515/184 TI Availability and Readiness Inspection of Severe Accident
Management Guidelines (SAMGs) (Section 4OA5.2)
2515/177 TI Managing Gas Accumulation in Emergency Core Cooling,
Decay Heat Removal, and Containment Spray Systems
(NRC Generic Letter 2008-01) (Section 4OA5.3)
Discussed
05000461/2009-005-00 LER Manual Scram on High Water Level Due to Reactor Recirc
[Recirculation] Pump Trip, Supplement 1 (Section 4OA3.1)
05000461/2010-002-05 FIN Failure to Correct Inadequate FWLCS Response Resulted in
High Reactor Vessel Water Level (Level 8) Scram
(Section 4OA3.1)
05000461/2008-001-00 LER Reactor Recirc [Recirculation] Pump Trip Initiates Automatic
Scram on High RPV [Reactor Pressure Vessel] Water Level
(Section 4OA3.2)
05000461/2008-001-01 LER Reactor Recirc [Recirculation] Pump Trip Initiates Automatic
Scram on High RPV [Reactor Pressure Vessel] Water Level
(Section 4OA3.2)
05000461/2008-004-01 FIN Failure to Perform Adequate Post-Maintenance Testing
Resulted in High Reactor Vessel Water Level (Level 8)
Scram (Section 4OA3.2)
2 Attachment
05000461/2008-004-02 FIN Failure to Evaluate an Unexpected and Unknown Cause for
Stray Voltage in the End-of-Cycle Recirculation Pump Trip
Circuit During Post-Modification Testing Resulted in a
Reactor Recirculation Pump Trip (Section 4OA3.2)
3 Attachment
LIST OF DOCUMENTS REVIEWED
The following is a list of documents reviewed during the inspection. Inclusion on this list does
not imply that the NRC inspectors reviewed the documents in their entirety, but rather, that
selected sections of portions of the documents were evaluated as part of the overall inspection
effort. Inclusion of a document on this list does not imply NRC acceptance of the document or
any part of it, unless this is stated in the body of the inspection report.
1R01 Adverse Weather Protection
- OP-AA-108-111-1001, Severe Weather and Natural Disaster Guidelines, Revision
- OP-AA-106-101-1002, Exelon Nuclear Issues Management,
- CPS 4302.01, Tornado/High Winds, Revision 19a
- AR 01204822, Entry Into CPS 4302.01, Tornado/High Winds Off-Normal
- AR 01204927, 345 kV South Bus Voltage Low Out of Band
- AR 01204929, Two Storm Related Events
- OP-AA-106-101-1002, Exelon Nuclear Issues Management, Revision 8
- OP-AA-108-111-1001, Severe Weather and Natural Disaster Guidelines, Revision 5
- CPS 4302.01, Tornado/High Winds, Revision 19a
- CPS 4303.02, Abnormal Lake Level, Revision 10
- CPS 4304.01, Flooding, Revision 5a
- WC-AA-107, Seasonal Readiness, Revision 9
- AC-CD-1105-0001, Elevated Lake Temps Challenge CPS Operating Parameters,
May 9, 2011
- Letter from F. Kearney to B. Hanson, Subject: Certification of 2011 Summer Readiness,
May 15, 2011
- M05-1059, P&ID Floor & Equip. Drains Screen House (DM), Sheet 3, Revision L
- A22-1032, Circulating Water Screen House Main Floor Plan Area-12 - El. 6990, Revision K
- AR 01216152, Initiate ACMP for Summer Operation (IR 1091600)
- AR 01210375, Initiate ACMP for Elevated Lake Temp in Prep for Summer
- AR 01091600, ACMP Needed for High Lake Temperature
- AR 01215817, NER, NC-011-012 Seasonal Readiness/Tornado
- AR 01200986, 345 KV 4520 B Phase Disconnect Elevated in Temperature
- AR 01076285, Elevated Temperature on 0SY4504C Disconnect Ball Side
- AR 01101328, Possible Vulnerability to a Summer Fish Loss at CPS
- AR 01092236, Gaps Identified During Effectiveness Review [of Industry Event Report
Recommendation Implementation]
- AR 01158006, Weakness in Implementation of [Industry Event Report Recommendation]
- AR 01207940, NOS ID: Material Staged in North Parking Lot Not Secured
- AR 01200727, Grid Transient Causes GCB [Gas Circuit Breaker] 4510 to Cycle Open and
Shut
- AR 01183058, 345 KV Switchyard Walkdown Issues/Results From 3/3/2011
- AR 01219519, 345 KV Switchyard Walkdown 5/23/11
- AR 01224933, ERAT SVC HVAC Unit #2 0VV90SB Found Not Running
- AR 01224407, 0VV89SA: RAT SVC Building HVAC Compressor Tripped
- AR 01224783, Unit 2 RAT B SVC Building HVAC Unit Not Providing Cooling
- AR 01073472, Work Order Chiller C Amps Cycling
- AR 01219150, 0Work Order02CE Low Oil Pressure
- AR 01219249, 0Work Order02CA: A Work Order Chiller Trip on Low Oil Pressure
- AR 01201962, Main Generator Issues Requiring Attention
4 Attachment
1R04 Equipment Alignment
- -AR 010294475, Procedure Enhancement
- -CPS 9082.01, Offsite Source Power Verification, Revision 39b
- -E02-1AP03, Electrical Loading Diagram Clinton Power Station Unit 1, Revision AA
- -CPS 3319.01, Standby Gas Treatment (VG), Revision 16
- -CPS 3319.01V001, Standby Gas Treatment Valve Lineup, Revision 8
- -CPS 3319.01V002, Standby Gas Treatment Instrument Valve Lineup, Revision 5a
- -CPS 3319.01E001, Standby Gas Treatment Electrical Lineup, Revision 10c
- -M05-1105, P&ID Standby Gas Treatment System (VG), Sheet 1, Revision S
- -M05-1105, P&ID Standby Gas Treatment System (VG), Sheet 2, Revision N
- -M05-1105, P&ID Standby Gas Treatment System (VG), Sheet 3, Revision F
- CPS 3402.01, Control Room HVAC (VC), Revision 25c
- CPS 3402.01E001, Control Room HVAC Electrical Lineup, Revision 10b
- CPS 3402.01V001, Control Room HVAC Valve Lineup, Revision 16e
- M05-1102, Control Room HVAC (VC), Revision U
- E02-OVC99, Schematic Diagram, Control Room HVAC System (VC), Revision R
1R05 Fire Protection
- CPS 1893.04M003, Prefire Plan Legend, Revision 1
- CPS 1893.04M625, 737 RadWaste: Paint & Oil Storage Room Prefire Plan, Revision 4
- CPS 1893.04M730, 777, 781, 783 Turbine: General Access and Mezzanines Prefire Plan,
Revision 5
- Calculation IP-M-0177, Fire Loads in Clinton Power Station
- Work Order 01347983, Secondary Containment Door 1DR1-263 Has Damaged Seal, June
21, 2010
- AR 00790021, Potable Water Valve Leakby Prevents Clearance Order Work
- AR 01075728, Secondary Containment Door 1DR1-263 Has Damaged Seal
- AR 01209630, NRC Observations/Questions in 737 Fuel Building
- Clinton Power Station Updated Final Safety Analysis Report, Appendix E, Fire Protection
Evaluation Report - Clinton Power Station Unit 1, Revision 11
- Clinton Power Station Updated Final Safety Analysis Report, Appendix F, Fire Protection Safe
Shutdown Analysis - Clinton Power Station Unit 1, Revision 11
- OP-AA-201-009, Control of Transient Combustible Material, Revision 11
- OP-CL-201-009, Control of Transient Combustible Material, Revision 1
- CPS 1893.04M410, 737 Fuel: Grade Level Prefire Plan, Revision 4a
- CPS 1893.04M622, 737 Radwaste: Drum Area and Bailer Room Prefire Plan, Revision 4
1R06 Flooding Protection Measures
- CPS 4304.01, Flooding, Revision 5a
- CPS Individual Plant Examination (IPE), Section 3.3.8, Internal Flood Analysis,
September 1992
- CPS 3219.01, CT [Containment], AB [Auxiliary Building], FB [Fuel Building] Floor Drain (RF),
Revision 8
- CPS-PSA-012, Clinton PRA 2003 Update Internal Flooding Update: Integration of the Internal
Flooding Analysis into the Single-Top Model, Revision 0
- CPS 4411.03, Injection/Flooding Sources, Revision 7
- CC-AA-309-1001, Suppression Pool Equalization Levels, Revision 5
- Clinton Power Station Updated Safety Analysis Report, Revision 13
5 Attachment
- NRC Information Notice 2009-006, Construction-Related Experiences with Flood Protection
Features, July 21, 2009
- Calculation 3C10-0485-001, Internal Flooding Analysis, Revision 8, Volume B
- SL-4576, Internal Flooding - Safe Shutdown Analysis and INPO SOER No. 85-5 Comparison
Evaluation Report (Sargent & Lundy), January 31, 1990
- A22-1032, Circulating Water Screen House Main Floor Plan Area-12 - El. 6990, Revision K
- AR 01197979, Flood Seals Do Not Have Periodic Inspection Program
- AR 01197992, Temporary Materials For Flood Mitigation Not Routinely Inventoried
- AR 01197991, Valve Used In Internal Flood Mitigation Not Accessible
- AR 01197988, Fuses Called Out In CPS 4304.01 Are Not Segregated
- AR 01197987, Hatches On SX Roof For Flood Access Procedure Weakness
- AR 01196294, NRC Senior Resident Identified Need To Improve Leak Berm
- AR 01092206, Functionality Review of Condenser Pit Level Switch
- AR 01023891, 1LSTF001B Failed to Actuate Per 3813.01
1R07 Heat Sink Performance
- ER-AA-340-1002, Service Water Heat Exchanger Inspection Guide, Revision 4
- CPS 2602.01, Heat Exchanger Performance of Shutdown Service Water Coolers Covered by
NRC Generic Letter 89-13, Revision 16b
- CPS 8130.01, Heat Exchanger Maintenance/Repairs, Revision 3
- Calculation IP-M-0486, SX Acceptance Flows/Area Reductions, Revision 6C
- Work Order 01238289, Inspect, Boroscope, Clean, Eddy Current, and Hydrolase as Required
1VX13AB Coil, April 6, 2011
- Drawing MC-136-415B, Nuclear Containment Cooling Coil 1VX13AA AB,
- Catalog ID 1150970, Coil, Cooling, Cleanable Tube Water, Left Hand, Half Serpentine, 3
Rows, April 6, 2011
- AR 01169271, Triennial Heat Sink and GL 89-13 FASA Deficiency, January 31, 2011
- AR 778875, 2700.12, Not Complete within 5Y Frequency, May 23, 2008
- AR00797796, HVAC Calculation Temperature Inconsistency, July 17, 2008
- AR01095477, Initial Results from Div 1 SX Flow Balance, July 28, 2010
- AR01210756, 1VY04S - RCIC Room Cooler Airflow Exceeds Max Allowed, May 2, 2011
- Calculation No. 024429, Formal Piping Stress Analysis for Shutdown Service Water
Subsystem 1SX-51, September 23, 2008
- CPS 1003.10, Clinton Power Station (CPS) Program for NRC Generic Letter 89-13 (Service
Water Problems Affecting Safety-Related Equipment), Revision 6d
- CPS 1938.04, Raw Water Vendor Interface Procedure, Revision 4d
- CPS 3209.01, Raw Water Treatment (RWT) System, Revision 18b
- CPS 3211.01, Shutdown Service Water (SX), Revision 25e
- CPS 4303.02, Abnormal Lake Level, Revision 10
- CPS 6069.01D001, SX System Operability Data Sheet, Revision 45a
- CPS5050.06, High/Low Temp RCIC Pump Room, Revision 35
- CR No. 1-99-02-367, Shutdown Service Water Divisions 1 and 2 Made Inoperable during
Cross Connected Operation, February 23, 1999
- CR No: 1-98-11-006; Operating Procedures May Inappropriately Change Plant
Configuration, November 2, 1998
- CY-AA-120-4110, Raw Water Chemistry Strategic Plan, Revision 6
- EC 264758; Replacement of Div 2 expansion Joints 1SX01MB, 1SX02MB, 1SX03MB,
1SX04MB with Carbon Steel Pipe, December 6, 2010
- EC 372365, Installation of Insulating Gaskets for Division 1 DG Heat Exchanger Expansion
Joints, October 22, 2008
6 Attachment
- ER-AA-340-1001, GL 89-13 Program Implementation Instructional Guide, Revision 7
- ER-AA-5400, Buried Piping and Raw Water Corrosion Program (BPRWCP) Guide,
- Revision 4
- IP-M-0734, Shutdown Service Water (SX) System Divisions I and II Hydraulic Transient Load
Evaluation, EC 349061 and 358900, Revision 0
- Performance Trend Data for 1SX01-PA, PB, and PC, April 1, 2011
- VC-86, Evaluation of Control Room Chillers for Shutdown Service Water System, Revision 1
- VY-01, VY System Cooling Load Calculation, Revision 9C
- VY-45, Performance Evaluation of VY System Cooling Coils Under SX Flow Acceptance
Limits, Revision 4E
- VZ-43, Maximum Water Flow for Cooling Coils and Refrigeration Condenser Served by WS
System, Revision 1B
- VZ-45, SX Room Cooler Airflow Test Evaluation, Revision 0B
- WO 594629, Perform DIV II System Testing IAW 2700.13, April 8, 2008
- WO00789724; Inspect, Boroscope, Clean, Eddy Current and Hydrolase 1VY04A; May 5, 2009
- WO00859883, Obtain Air Flow Measurements for Room Cooler Coils 1VY04S,
August 9, 2007
- WO00864362, Major Inspection; Hydrolaze 0VC13CA Chiller, May 29, 2008
- WO01057470, Inspection/Clean Condenser; Hydrolance Tubes 0VC13CB Chiller,
November 12, 2008
- WO01195711, Div I SX System Testing IAW 2700.12, July 28, 2010
- WO01284960, Clean and Inspect 1VY04A Coil, May 3, 2011
1R12 Maintenance Effectiveness
- Regulatory Guide 1.160, Monitoring the Effectiveness of Maintenance at Nuclear Power
Plants, Revision 2 March 1997
- NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at
Nuclear Power Plants, Revision 2
- ER-AA-310, Implementation of Maintenance Rule, Revision 8
- ER-AA-310-1001, Maintenance Rule Scoping, Revision 4
- ER-AA-310-1005, Maintenance Rule - Dispositioning Between A(1) and A(2), Revision 5
- AR 00944238, 2)
- Clinton Power Station Updated Safety Analysis Report, Revision 13
- Regulatory Guide 1.160, Monitoring the Effectiveness of Maintenance at Nuclear Power
Plants, Revision 2 March 1997
- NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at
Nuclear Power Plants, Revision 2
- ER-AA-310, Implementation of Maintenance Rule, Revision 8
- ER-AA-310-1001, Maintenance Rule Scoping, Revision 4
- Maintenance Rule Scoping and Performance Criteria for Radiation Monitoring System,
May 23, 2011
- Common Cause Analysis 01179979, Potential Trend on Radiation Monitor Failures,
March 24, 2011
- AR 01179979, Potential Trend on Radiation Monitor Failures
- AR 01093695, 1RIX-PR042C Failed High
- AR 01090862, 1RIX-PR042B Failed Downscale to 0 Mr/Hr
- AR 01165170, 0RIX-PR001 High Range Noble Gas Channel Failed Calibration
- AR 01179569, 0RIX-PR001 Sample Pump Failed
- AR 01178074, 1RIX-PR006A High Alarm, Spike
7 Attachment
- AR 01204337, Received Unexpected AR/PR Hi Alarms on 1RIX-PR006A
- AR 01205697, 0RIX-PR004, SGTS Radiation Monitor in Communications Failure
1R13 Maintenance Risk Assessments and Emergent Work Control
- CPS 3310.01, Reactor Core Isolation Cooling (RI), Revision 27D
- CPS 9054.01, RCIC System Operability Check, Revision 43
- CPS 9054.01C001, RCIC Water Leg Pump 1E51-C003 Operability Test 1E51-F040 Closure
Test and 1SX037 Stroke Timing, Revision 6B
- CPS 9054.01C002, RCIC 1E51-C001 High Pressure Operability Checks, Revision 3A
- CPS 9054.02, RCIC Valve Operability Checks, Revision 38C
- Work Order 01170713-01, Replace EG-M Box Every 8 yrs, May 6, 2011
- Work Order 01278032-01, Inspect Suppression Pool Suction Check Valve, May 19, 2011
- AR 01211183, WW 1119 SOW Logic Bust With IMD Work
- AR 01211506, 1E51C003: C/O Required For PM, But Not Requested
- AR 01211665, Water from Vent During RCIC SOW
- AR 01212752, WW 1119 RI SOW Unavailable Hours 119% of Scheduled
- ER-AA-600, Risk Management, Revision 6
- ER-AA-600-1012, Risk Management Documentation, Revision 9
- ER-AA-600-1042, On-Line Risk Management, Revision 7
- WC-AA-101, On-Line Work Control Process, Revision 18
- WC-AA-104, Integrated Risk Management, Revision 18
- Clinton Power Station Technical Specifications
- AR 01091836, Plant Risk Yellow Entered When Not Required
1R15 Operability Evaluations
- ER-AA-2009, Managing Gas Accumulation, Revision 1
- Operability Evaluation 384223, 1E21-F303 Leaking By Seat, Revision 0
- Illinois Power Condition Report 1-95-09-025, Check Valve Failure, Revision 0
- Maintenance Request Number D50975, Rework Check Valve to Restore its Function
- Maintenance Request Number D61556, LPCS Test Line Check Valve
- CPS 1401.09F002, Cat A Instrument Failure Checklist, Revision 1
- CPS 9052.01, LPCS/RHR A Pump & LPCS/RHR A Water Leg Pump Operability,
Revision 46a
- CPS 9052.01D001, LPCS/RHR A Pump & LPCS/RHR A Water Leg Pump Operability Data
Sheet, Revision 43d
- CPS 9082.02, Electrical Distribution Verification, Revision 35c
- AR 01204102, Cat A Failure of HPCS Instrument
- AR 01208215, 1E21F303 Abnormal Flow/Indication During LPCS Clearance Hang
- AR 01208296, 1401.09 Enhancements to Cat A Instrument Failure Process
- Work Order 01336205-01, 9052.01 LPCS Pump Operability, July 29, 2010
- M05-1073, Low-Pressure Core Spray (LPCS)(LP), Sheet 1, Revision AG
- M05-1075, Residual Heat Removal (RH), Sheet 1, Revision AW
- M05-1075, Residual Heat Removal (RH), Sheet 4, Revision AF
- Clinton Power Station Technical Specifications
- Clinton Power Station Updated Final Safety Analysis Report, Revision 13
- NRC Regulatory Issue Summary 2005-20, Revision to NRC Inspection Manual Part 9900
Technical Guidance, Operability Determinations & Functionality Assessments for Resolution
of Degraded or Nonconforming Conditions Adverse to Quality or Safety, Revision 1
- EC 384575, High Vibration Levels on 0VC04CB, Revision 0
8 Attachment
- EC 384092, NRC Question on RCS PIV Surveillance Testing, Revision 0
- AR 01219600, Vibration Levels Increased on 0VC04CB
- AR 01202456, NRC Question on RCS PIV Surveillance Testing
- AR 01194749, Division 1 DG Slow Start
- AR 01194803, Transient Test Servers Full - Impact DG Surveillance
- CPS 9080.24, DG 1A Test Mode Override, Load Reject Operability, and Idle Speed
Override, Revision 3a
1R18 Plant Modifications
- AR 01121419, 1DG01KA-A10/A11 DIV I DG Overvoltage Breaker Tripped
- EC 381638, Temporary Modification to Lift Input from A10 Device to A11 Device for the
Division I Diesel Governor, Revision 0
- E02-1DG99, Schematic Diagram Diesel Generator 1A Excitation, Sheet 016, Revision M
- 50.59 Screening Number CL-2010-S-029 for EC #381638, Temporary Modification to Lift
Input from A10 Device to A11 Device for the Division I Diesel Governor, Revision 0
1R19 Post-Maintenance Testing
- EC 379884, Replace Gould Type J13 Auxiliary Relays 1UAY-DG292 and KL With GE
CR120BD Relay in Division 2 EDG Control Panel 1PL12JB, Revision 0
- ECR 394681, DC Operated Type J13 Auxiliary Relay, Revision 0
- CPS 3310.01, Reactor Core Isolation Cooling (RI), Revision 27d
- CPS 9054.01, RCIC System Operability Check, Revision 43
- CPS 9054.01C002, RCIC (1E51-C001) High Pressure Operability Checks, Revision 3a
- CPS 9054.01D002, RCIC (1E51-C001) ) High Pressure Operability Check, Revision 43
- CPS 9054.02, Reactor Core Isolation Cooling Valve Operability Checks Checklist,
Revision 23f
- CPS 9054.02D001, RCIC Valve Operability Data Sheet, Revision 39d
- CPS 9070.01, Control Room HVAC Air Filter Package Operability Test Run, Revision 26d
- CPS 9080.19, DG 1B Overcrank Delay Timer Test, Differential Overcurrent Trip Test, and
Trip Bypass Operability, Revision 0c
- CPS 9080.19D001, DG 1B Overcrank Delay Timer Test Data Sheet, Revision 0
- Work Order 01278032-02, OP PMT for 1E51F030, May 5, 2011
- Work Order 01283985-03, OP PMT for 0VC03CA, April 27, 2011
- Work Order 01294892-01, 1PL12JB: Replace the Division 2 EDG LOCA Bypass Relay
3KL4, March 1, 2011
- Work Order 01294892-03, OPS PMT 9080.19, April 13, 2011
- Work Order 01298883-01, 9054.02D20 OP RCIC Valve Operability (1E51-F079, 81 Only),
May 5, 2011
- Work Order 01363959-02, Generic Replacement of Safety Related Love Controllers,
April 28, 2011
- Work Order 01411430-01, OP 9054.02 RCIC Valve Operability, May 5, 2011
- Work Order 01413537-01, OP 9054.01C002 RCIC 1E51-C001 High Pressure Operability
Check, May 6, 2011
- Work Order 01176252-01, Test Bus 1A1 Main Feed Breaker Protective Relays
- Work Order 00918051-03, Replace Rosemount 1153 Transmitter
- Work Order 01175527-02, Replace and Calibrate Capacity Controller 1TCVP013
- Work Order 01175621-07, PMT for 1TSVP085A & 1TE-VP013
- AR 00985349, 1DG01KA: EDG Div 1 Did Not Go To Full Speed When In Run
- AR 00985660, Found Relay 1UAYDG291 Bad While Troubleshooting 1DG01KA
9 Attachment
- AR 01208618, VC-A SOW PMT Logic Incorrect Delaying Restoration
- AR 01212052, 1E51-F030: Unexpected Torque Readings During Valve Stroke
- AR 01212058, During Performance of 9054.02 Unexpected Values Obtained
- AR 01212267, As Found Torque Is LOOS
- AR 01212373, LL - 9054.02 RCIC Check Valve Operability 1E51-F079
- AR 01212535, 1E51F079: Damaged Set Screw
- AR 01213246, Improved Safety While Testing the 1E51-F030
- M05-1079, RCIC, Sheet 1 Revision AH
- M05-1079, RCIC, Sheet 2 Revision AJ
- Work Order 01405272-02, "0FP03P Outboard Bearing Elevated Temperature"
- Work Order 01404960-01, Horizontal Fire Pump: Perform Operability Test IAW CPS 3822.01
- CPS 3822.06, Operation of the Horizontal Fire Pump, Revision 9
- AR 01203214, Horizontal Fire Pump Packing Hot During Maintenance Run
- AR 01198630, 0FP003 Fire Pump Discharge Packing Leakage Increased
- AR 01198618, 0FP03P (Horizontal FP) INDB/OUTBD Packing Smoking
- AR 01205147, Charger B Voltmeter Failed High 30 VDC (0FP03P)
- AR 01209830, 0FP03P Positive Battery Connector Found Broken on Batter #2
1R22 Surveillance Testing
- CPS 9051.01, HPCS Pump & HPCS Water Leg Pump Operability, Revision 44
- CPS 9051.01D001, HPCS Pump & HPCS Water Leg Pump Operability Data Sheet,
Revision 45
- CPS 9051.02, HPCS Valve Operability Test, Revision 40b
- CPS 9053.04, Residual Heat Removal (RHR) A/B/C Valve Operability Checks, Revision 45b
- CPS 9053.04C002, RHR Loop B Valve Operability, Revision 1b
- CPS 9053.04D002, RHR Loop B Valve Operability Data Sheet, Revision 34b
- CPS 9080.01, Diesel Generator 1A Operability - Manual and Quick Start Operability,
Revision 52.e
- CPS 9866.01, VG/VC [Standby Gas Treatment/Control Room Ventilation] HEPA [High
Efficiency Particulate Air] Filter Leak Test, Revision 26
- Work Order 01403904-01, OP 9051.02 HPCS Valve Operability (Stroke Time), April 18, 2011
- Work Order 01420090-01, Op Perform RHR B Valve Operability Per 9053.04C002, June 13,
2011
- AR 01204162, HPCS Surveillance Enhancement 9051.02 and .05
- Clinton Power Station Technical Specifications
- Clinton Power Station Updated Final Safety Analysis Report, Revision 13
- Clinton Nuclear Power Station Unit 1, Inservice Testing Program Plan - Third Ten Year
Interval, Revision 0
- HU-AA-1212, Technical Task Risk/Rigor Assessment, Pre-job Brief, Independent Third Party
Review, and Post-job Review, Revision 4
- Apparent Cause Evaluation AR 01212825, NRC URI 2011002-04: RCS PIV Leakage
Surveillance Test, June 21, 2011
- CPS 9843.01, ISI [Inservice Inspection] Category A Valve Leak Rate Test, Revision 35
- CL-SURV-10, Risk Analysis for Potentially Deficient Surveillance High to Low Pressure
Interface Valves May Have Been Tested Using Too High a Differential Pressure, Revision 0
- Work Order 1144785-01, MC010-1 LLRT [Local Leak Rate Test] FW [Feedwater] B Line
9861.05D014, January 19, 2010
- Work Order 1144795-01, MC009-1 LLRT FW A Line 9861.05D013, January 17, 2010
- Work Order 1144801-01, 9843.01V003 Category A Valve Leak Rate Test (1E21-F005) LPCS
Injection, January 13, 2010
10 Attachment
- Work Order 1144802-01, 9843.01V003 Category A Valve Leak Rate Test (1E21-F006) LPCS
Injection, January 13, 2010
- Work Order 00790605-01, 1E21-F006 Contingent Rework on LLRT Failure, January 16,
2010
- Work Order 1144820-01, 9843.01V004 Category A Valve Leak Rate Test (1E12-F042C) LPCI
[Low Pressure Coolant Injection] C Drywell Isolation, January 19, 2010
- Work Order 1144817-01, 9843.01V004 Category A Valve Leak Rate Test (1E12-F041C) LPCI
C Test Check Valve, January 19, 2010
- Work Order 1144792-01, 9843.01V018 Category A Valve Leak Rate Test (1E12-F499A/B,
497) RHR Keep Fill, January 19, 2010
- Work Order 1144814-01, 9843.01V001 Category A Valve Leak Rate Test (1E12-F041A) LPCI
A Test Check Valve, January 13, 2010
- Work Order 1144818-01, 9843.01V001 Category A Valve Leak Rate Test (1E12-F042A) LPCI
A Drywell Isolation, January 13, 2010
- Work Order 1144793-01, 9843.01V019 Category A Valve Leak Rate Test (1E12-F495A/B,
496) RHR Keep Fill, January 14, 2010
- Work Order 1144819-01, 9843.01V003 Category A Valve Leak Rate Test (1E12-F042B) LPCI
B Drywell Isolation, January 21, 2010
- Work Order 1144815-01, 9843.01V003 Category A Valve Leak Rate Test (1E12-F041B) LPCI
B Test Check Valve, January 21, 2010
- Work Order 1144828-01, 9843.01V015 Category A Valve Leak Rate Test (1E51-F066) RCIC
Header Spray, January 17, 2010
- Work Order 1144796-01, 9843.01V005 Category A Valve Leak Rate Test (1E22-F004) HPCS
Injection, January 15, 2010
- Work Order 1144797-01, 9843.01V005 Category A Valve Leak Rate Test (1E22-F005) HPCS
Injection, January 15, 2010
- Work Order 1141926-01, 1E22F005 Contingent Repair in Event of LLRT Failure, January 21,
2010
- Work Order 1144810-01, 9843.01V006 Category A Valve Leak Rate Test (1E12-F008) RHR
Shutdown Cooling Suction, January 17, 2010
- Work Order 1144812-01, 9843.01V006 Category A Valve Leak Rate Test (1E12-F009) RHR
Shutdown Cooling Suction, January 17, 2010
- Work Order 1144822-01, 9843.01V009 Category A Valve Leak Rate Test (1E12-F023)
Reactor Pressure Vessel Head Spray, February 1, 2010
- Work Order 1144826-01, 9843.01V009 Category A Valve Leak Rate Test (1E51-F059) RCIC
Test Return to RCIC Storage Tank, February 1, 2010
- AR 01016798, 9843.01D002 Error in Corrected Pressure Calculation
- AR 01198669, Senior Resident NRC Inspector Noted Deficiencies in C1R12 Leak Rate
Testing
- AR 00282084, Discrepancy Between TSSR 3.4.6 and CPS 9843.01
- AR 01202456, NRC Question on RCS PIV Surveillance Testing
- AR 01212825, NRC URI 2011002-04: RCS PIV Leakage Surveillance Test
- AR 01207467, Potential Creep Away From Meeting Regulatory Requirements
- AR 01239007, NRC Identified - VC Flow Issue
2RS3 In-Plant Airborne Radioactivity Control and Mitigation
- RP-AA-302, Determination of Alpha Levels and Monitoring, Revision 3
- RP-AA-440, Respiratory Protection Program, Revision 9
- RP-AA-825, Maintenance, Care and Inspection of Respiratory Protective Equipment,
- Revision 3
11 Attachment
- ProCheck3 Test Results, MSA MMR/Firehawk 4500, Serial Number OY241621, May 6, 2010
- ProCheck3 Test Results, MSA MMR/Firehawk 4500, Serial Number OY241621, June 26,
2009
- ProCheck3 Test Results, MSA MMR/Firehawk 4500, Serial Number QY130249, July 9, 2009
- ProCheck3 Test Results, MSA MMR/Firehawk 4500, Serial Number QY130249,
July 11, 2010
- ProCheck3 Test Results, MSA MMR/Firehawk 4500, Serial Number QY130253, July 9, 2009
- ProCheck3 Test Results, MSA MMR/Firehawk 4500, Serial Number QY130253,
July 11, 2010
- RP-CL-825-101, CPS Maintenance and Care of Respiratory Protective Equipment,
- Revision 33
- System Walkdown for VC (Control Room Ventilation), December 28, 2010
- System Walkdown for VG (Standby Gas Treatment), December 28, 2010
- Work Order Package 01079296 02, 9866.1 Perform HEPA Filter Test on 0VC09SA,
- April 25, 2009
- Work Order Package 00811833 05, Perform HEPA Filter Test on 0VC09SA per CPS 9866.1,
May 2, 2077
- Work Order Package 0109925 05, 9866.1 Perform HEPA Filter Test on 0VC09SB, July 14,
2009
- Work Order Package 00833431 05, Perform HEPA Filter Test on 0VC09SB per CPS 9866.1,
July 17, 2007
- Work Order Package 01136958 05, 9866.1 Perform HEPA Filter Test on 0VG07FB and
0VGG11FB, March 8, 2010
- Work Order Package 01136958 05, 9866.1 Perform HEPA Filter Test on 0VG07FB and
0VGG11FB, March 8, 2010
- Work Order Package 00842268 06, Perform HEPA Filter Test on 0VG07FB and 0VGG11FB
per CPS 9866, November 13, 2007
- Work Order Package 01293106 02, 9866.1 Perform HEPA Filter Test on 0VG07FA and
0VGG11FA, March 1, 2011
- Work Order Package 01089496 025, 9866.1 Perform HEPA Filter Test on 0VG07FA and
0VGG11FA, June 2, 2009
- Work Order Package 01079297 01, 9866.02 Perform Charcoal Adsorber Leak Test
0VC7SA/9SA, April 28, 2009
- Work Order Package 00811832 01, Perform Charcoal Adsorber Leak Test 0VC7SA/9SA per
CPS 9866, May 5, 2007
- Work Order Package 01099926 01, 9866.02 Perform Charcoal Adsorber Leak Test
0VC7SB/9SB, July 7, 2009
- Work Order Package 01265331 02, 9866.02 Perform Charcoal Adsorber Leak Test
0VG08FA, March 1, 2011
- Work Order Package 01089817 02, 9866.02 Perform Charcoal Adsorber Leak Test
0VG08FA, March 2, 2009
- Work Order Package 01136959 05, 9866.02 Perform Charcoal Adsorber Leak Test
0VG08FB, March 18, 2010
- Quarterly Service Air and Self Contained Breathing Apparatus, December 17, 2009
- Quarterly Service Air and Self Contained Breathing Apparatus, March 12, 2010
- Quarterly Service Air and Self Contained Breathing Apparatus, June 25, 2010
- Quarterly Service Air and Self Contained Breathing Apparatus, July 30, 2010
- Quarterly Service Air and Self Contained Breathing Apparatus, September 17, 2010
- Quarterly Service Air and Self Contained Breathing Apparatus, February 23, 2011
- AR 01214577, Mask in Premaire Unit Found with Bad Exhalation Diaphragm, May 11, 2011
- AR 01215101, Storage of Licensed Operator Respirator Spectacle Kits, May 12, 2011
12 Attachment
- AR 01215230, Review Need for Validation of Respirator Spectacle Kits, May 12, 2011
- AR 01215184, SCBA HUD Batteries are not the Recommended Batteries, May 12, 2011
- AR 0121513, SCBA Face piece Drying Gap, May 12, 2011
- MSA MMR Air Mask with Firehawk Regulator, Operating and Instruction Manual, TAL 406
(L), Revision 12
2RS4 Occupational Dose Assessment
- NUPIC Audit SA10-017, Mirion Technologies (GDS) Inc., January 3, 2011
- RP-AA-210-2001, Ability to Wear the Thermoluminescent Dosimeter (TLD) Under Protective
Clothing, Revision 0
- RP-AA-11, External Dose Control Program, Revision 0
- RP-AA-12, Internal Dose Control Program Description, Revision 1
- RP-AA-203; Exposure Control and Authorization, Revision 3
- RP-AA-203-1001, Personnel Exposure, Revision 6
- RP-AA-210, Dosimetry Issue, Usage, and Control, Revision 20
- RP-AA-210-1001, Dosimetry Logs and Forms, Revision 5
- RP-AA-214, Area TLD Surveillance, Revision 3
- RP-AA-220 Bioassay Program, Revision 7
- RP-AA-221, Whole Body Data Review, Revision 1
- RP-AA-222, Methods for Estimating Internal Exposure from In Vivo and In Vitro Bioassay
Data, Revision 3
- RP-AA-230, Operation of the Canberra FASTSCAN Whole Body Counter, Revision 0
- RP-AA-250, External Dose Assessments from Contamination, Revision 5
- Calibration of the Canberra FASTSCAN WBC System at the Clinton Power Station, 2/24/2011
- Audit SA 10-017; QAD2011001, Joint Audit of Mirion Technologies (GDS) Inc., January 3,
2011
- Audit SR 2008-001, Joint Audit of Global Dosimetry Solutions, January 10, 2008
- FASA, Occupational Dose Assessment & In Plant Airborne Radioactivity Control & Mitigation,
Assignment 1056527-03, February 1, 2011
- RP-AA-270, Prenatal Radiation Exposure, Revision 6
- AR 01215225, Passport Expiration Date of DPW Needs Improvement, May 12, 2011
- AR 01215180, RP Procedure Enhancement to Clarify Attachment Use, May 12, 2011
4OA1 Performance Indicator Verification
- AR 01113608, Div 2 EDG Quick Start Time > 9080.02 Step 9.1.6 Criteria
- AR 01187358, Change 9080.26 to Eliminate an Unnecessary Engine Start
- AR 01194749, Division 1 Slow Start Time
- AR 01214578, 1DG01KB: D2 DG Tripped During 9080.02
- CPS 9000.01D001, Control Room Surveillance Log - Mode 1, 2, 3 Data Sheet, Revision 52e
- LS-AA-2200, Mitigating System Performance Index Data Acquisition & Reporting, Revision 3
- LS-AA-2001, Collecting and Reporting of NRC Performance Indicator Data, Revision 14
- Nuclear Energy Institute 99-02, Regulatory Assessment Performance Indicator Guideline,
Revision 6
- RM Document Number CL-MSPI-01, Clinton MSPI Basis Document, Revision 5
- MSPI Derivation Reports, Period March 2011, for Emergency AC Power System
- Nuclear Energy Institute 99-02, Regulatory Assessment Performance Indicator Guideline,
Revision 6
13 Attachment
4OA2 Identification and Resolution of Problems
- ANSI/ANS 56.8-2002, "Containment System Leakage Testing Requirements"
- NEI 94-01, "Industry Guideline for Implementing Performance-based Option of 10 CFR Part
50, Appendix J"
- NRC Information Notice 85-71, Containment Integrated Leak Rate Tests
- Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program"
- CPS 1305.01, Primary Containment Leakage Rate Testing Program, Revision 10c
- CPS 1305.01F001, Type 'B' Local Leak Rate Summary Sheet, Revision 2
- CPS 9861.04, MSIV Local Leak Rate Test (MC-5,6,7,8), Revision 26
- CPS 9861.04D002, MSIV B Local Leak Rate Test Data Sheet (1MC-8), Revision 25d
- Work Order 01128244, MC008 LLRT Requirements (MSIV - B) and PIT 1E32-F001E,
January 20, 2010
- Operational and Technical Decision Making (OTDM) #1229710, Through Wall Steam Leak on
1MS13AA-2
- RCR 1021241, Late Identification of Work Scope for 1B21F022C, Inboard Main Steam Line C
Isolation Valve
- EACE 1017464, Investigate Failure of 'B' MSIVs
- AR 01017464, 1B21F028A: 9861.04 LLRT on MSL A, B, and C Test Failure
- AR 01059673, NOS ID MSIV As-Found Results Re-Evaluate Reportability
- AR 01224527, NRC PI&R: As-Found LRT For Each MSIV Not Performed In C1R12
- AR 01228126, Heater Bay Hotter Than Expected
- AR 01229320, Steam Leak Identified On 1ES001B
- AR 01229325, 1WO03SL - Water Dripping Near 1FW01AA 6A HP Heater
- AR 01229569, 1ES001A Has Small Packing Leak
- AR 01229710, Through Wall Steam Leak On 1MS13AA-2
- AR 01231642, Need Contingent Actions For High Heater Bay Temperatures
- AR 01232761, Water Flow Check For Turbine Building Area Coolers
- AR 01233539, Replace 2 Inch Pipe 1MS13AB Downstream Of Valve 1B21CA6
- AR 01233540, Replace 2 Inch Pipe 1MS13AC Downstream of Valve 1B21CA5
4OA3 Followup of Events and Notices of Enforcement Discretion
- LER 05000461/2008-001-02, "Reactor Recirculation Pump Trip Initiates Automatic Scram on
High RPV Water Level," Supplement 2
- LER 05000461/2009-005-01, "Manual Scram on High Water Level Due to Reactor
Recirculation Pump Trip," Supplement 1
4OA5 Other Activities
- 0000-0088-8669-R0, BWR Owners Group Technical Report Effects of Voiding in ECCS
Drywell Injection Piping, September 2008
- 0000-0088-8669-R0, BWR Owners Group Technical Report; Effects of Voiding in ECCS
Drywell Injection Piping, September 2008
- 3C10-0175-001, Design and Analysis of Clinton Containment Spray System, Revision 3A
- AR 00807753, NRC GL 08-01 Inspection Results At Pipe 1RH50AB
- AR 00812163, NRC GL 2008-01 Inspection Results at Pipe 1RH
- AR 01212387, NRC GL 2008-01 Lack of Gas Management RHR Discharge Piping Void
- AR00802940, GL 08-01 Inspection Results at 1E12F037A, August 1, 2008
- AR00807753, GL 08-01 Inspection Results at 1RH50AB, August 15, 2008
- AR00812163, GL 08-01 Inspection Results at 1RH03AA, August 28, 2008
14 Attachment
- AR00814512, GL 08-01 Inspection Results at 1RH117A, September 5, 2008
- AR01022886, RHR C Pump Suction Voiding, January 28, 2010
- AR01173402, FASA Eval Adding Time Duration to Venting Act, February 10, 2011
- AR01195401, GL 2008-01 Inspection Findings at Byron/Braidwood, March 31, 2011
- AR01195408, GL 2008-01 Inspection Findings at Byron/Braidwood, March 31, 2011
- AR01197314, GL 2008-01 Inspection Findings at Byron/Braidwood, April 4, 2011
- AR01212387, Lack of Gas Management RHR Discharge Piping Void, May 5, 2011
- ATI-992573-07, NRC IN 2010-11 Voiding in RHR Piping
- CPS 3309.01, High Pressure Core Spray (HPCS), Revision 16a
- CPS 3312.01, Residual Heat Removal, Revision 38c
- CPS 3312.03, RHR - Shutdown Cooling (SDC) and Fuel Pool Cooling and Assist (FPC&A),
Revision 6c
- CPS 9051.01, HPCS Pump and HPCS Water Leg Pump Operability, Revision 44a
- CPS 9051.05, HPCS Discharge Header Filled and Flow Path Verification, Revision 27e
- CPS 9053.01, RHR B/RHR C Discharge Header Filled and Flow Path Verification,
Revision 28F
- EC 371529, Generic Letter 2008-01 HPCS Evaluation, Revision 1
- EC 371531, GL 2008-01 System Evaluation Template, Exelon Specific, Clinton Power Station
- RHR Evaluation, Revision 1
- EC 371609, Ultra Sonic Inspection Criteria: Division 1 ECCS: RHR A/LPCS, Revision 1
- EC 371659, Generic Letter 2008-01 Air Intrusion in ECCS Systems Ultrasonic Inspection
Criteria Division 2 ECCS: RHR-B/RHR-C; Revision 1,
- EC 371983, Installation of High Point vent on Line 1RH50AB-10 Cancel to EC373186 and
Calc IP-M-0777, August 16, 2010
- EC 373186, Piping Air Pocket acceptance (NRC GL 2008-01), Valve Bonnet and Known
Pockets, Revision 0
- EC-371560, HPCS Vent Modification, Revision 0
- EC-371660, Generic Letter 2008-01: Air Intrusion in ECCS Systems Ultra-sonic Inspection
Criteria: Division 3 ECCS: HPCS, Revision 1
- EC-380824, Generic Letter 2008-01 System Periodic UT Frequency Evaluation Clinton Power
Station - RHR, LPCS and HPCS; Revision 0
- ER-AA-2009, Managing Gas Accumulation, Revision 1
- ER-AA-335-007, Ultrasonic Inspection for Determination of Sedimentation in Piping Systems
or Components and Fluid Level Measurements, Revision 3
- FAI/08-70, Gas Void Pressure Pulsations Program, Revision 0
- HP-1, High Pressure Core Spray Isometric Drawing;Revision 7U
- HP-2, High Pressure Core Spray Isometric Drawing; Revision 10L
- HP-3, High Pressure Core Spray Isometric Drawing; Revision 6A
- HP-4, High Pressure Core Spray Isometric Drawing; Revision 7E
- HP-5, High Pressure Core Spray Isometric Drawing; Revision 9N
- HP-6, High Pressure Core Spray Isometric Drawing; Revision 6R
- M05-1074, P&ID High Pressure Core Spray; Revision AH
- M05-1075-001, P& ID Residual Heat Removal (RH), Revision AW
- M05-1075-002, P& ID Residual Heat Removal (RH), Revision AM
- M05-1075-003, P& ID Residual Heat Removal (RH), Revision AG
- Operability Evaluation 812163-02, Residual Heat Removal System, January 13, 2009
- Power Point Presentation on Training for GL 2008-01
- RH-09, System: Residual Heat Removal Isometric Drawing, Revision 7A
- RH-11, System: Residual Heat Removal Isometric Drawing, Revision 5M
- RH-14, System: Residual Heat Removal Isometric Drawing, Revision 12M
- RH-17, System: Residual Heat Removal Isometric Drawing, Revision 8H
15 Attachment
- RH-21, System: Residual Heat Removal Isometric Drawing, Revision 11L
- RS-08-131, Nine-Month Response to Generic Letter 2008-01, October 14, 2008
- RS-09-173, Response to Request for Additional Information Regarding Generic Letter 2008-01, December 15, 2009
- WO01359924, UT Testing to Check for Accumulated Air - HPCS, October 18, 2008
- WO01379900, UT Testing to Check for Accumulated Air - HPCS, January 17, 2011
- AR01206227, Missed Impacts to RCIC Cooling Load Calculations, April 22, 2011
- AR01208619, 12-Minutes Basis in Procedure - Injection Piping Fill Time, April 27, 2011
- AR01212387, NRC GL 2008-01 Lack of Gas Management RHR Discharge Piping Void,
May 5, 2011
- AR01205245, SX Rooms Watertight Doors SD1-11 and SD1-12 are Found Open,
April 20, 2011
- AR01205404, Document Update Missed in EC, April 20, 2011
- AR01209715, NRC IN 2010-11 Response, April 29, 2011
16 Attachment
LIST OF ACRONYMS USED
AC Alternating Current
ADAMS Agency-wide Documents and Management System
AR Action Request
CFM Cubic Feet Per Minute
CFR Code of Federal Regulations
CNO Chief Nuclear Officer
DG Diesel Generator
ECCS Emergency Core Cooling System
°F Degrees Fahrenheit
FIN Finding
GL Generic Letter
GPM Gallons Per Minute
HEPA High Efficiency Particulate Air
HVAC Heating Ventilation and Air Conditioning
IMC Inspection Manual Chapter
IP Inspection Procedure
ISI Inservice Inspection
LER Licensee Event Report
LLRT Local Leak Rate Test
LOCA Loss-of-Coolant-Accident
LPCS Low Pressure Core Spray
MSPI Mitigating Systems Performance Index
NCV Non-Cited Violation
NEI Nuclear Energy Institute
NIOSH National Institute for Occupational Safety and Health
NRC U.S. Nuclear Regulatory Commission
NRR Office of Nuclear Reactor Regulation
P&ID Piping and Instrumentation Diagram
PARS Publicly Available Records System
% Percent
PIV Pressure Isolation Valve
PMT Post-Maintenance Test
PSIG Pounds Per Square Inch Gauge
RCIC Reactor Core Isolation Cooling
SAMG Severe Accident Management Guidelines
SBO Station Blackout
SCBA Self Contained Breathing Apparatus
SDP Significant Determination Process
SSC Structures, System, and Component
SX Shutdown Service Water
TEDE Total Effective Dose Equivalent
17 Attachment
TI Temporary Instruction
TR Topical Report
TS Technical Specification
TSO Transmission System Operator
TSSR Technical Specification Surveillance Requirement
TSTF Technical Specification Task Force
UFSAR Updated Final Safety Analysis Report
URI Unresolved Item
VC Control Room Ventilation
VG Standby Gas Treatment
Work Order Work Order
WR Work Request 18 Attachment
M. Pacilio -2-
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter,
its enclosure, and your response (if any) will be available electronically for public inspection in
the NRC Public Document Room or from the Publicly Available Records System (PARS)
component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Website
at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Mark A. Ring, Chief
Branch 1
Division of Reactor Projects
Docket No. 50-461
License No. NPF-62
Enclosure: Inspection Report 05000461/2011-003
w/Attachment: Supplemental Information
cc w/encl: Distribution via ListServ
DISTRIBUTION:
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To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl
"E" = Copy with attach/encl "N" = No copy
OFFICE Clinton RIO RIII E RIII RIII
NAME MRing for BKemker MRing:cs
DATE 07/29/11 07/29/11
OFFICIAL RECORD COPY
Letter to M. Pacilio from M. Ring dated July 29, 2011
SUBJECT: CLINTON POWER STATION, NRC INTEGRATED INSPECTION REPORT
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