ML12031A265
| ML12031A265 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 01/31/2012 |
| From: | Ring M NRC/RGN-III/DRP/B1 |
| To: | Wells P NextEra Energy Duane Arnold |
| References | |
| IR-11-005 | |
| Download: ML12031A265 (45) | |
See also: IR 05000331/2011005
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION III
2443 WARRENVILLE ROAD, SUITE 210
LISLE, IL 60532-4352
January 31, 2012
Mr. Peter Wells
Vice President
NextEra Energy Duane Arnold, LLC
3277 DAEC Road
Palo, IA 52324-9785
SUBJECT:
DUANE ARNOLD ENERGY CENTER - NRC INTEGRATED INSPECTION
REPORT 05000331/2011005
Dear Mr. Wells:
On December 31, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed an
inspection at your Duane Arnold Energy Center. The enclosed inspection report documents the
inspection results which were discussed on January 12, 2012, with you and other members of
your staff.
The 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.
One NRC-identified traditional enforcement Severity Level IV violation and two
self-revealed findings of very low safety significance (Green) were identified during this
inspection. The two findings were determined to also involve violations of NRC requirements.
Further, a licensee-identified violation, which was determined to be of very low safety
significance, is listed in this report. The NRC is treating these violations as non-cited violations
(NCVs) consistent with Section 2.3.2 of the NRC Enforcement Policy.
If you contest these non-cited violations, you should provide a response within 30 days of the
date of this inspection report, with the basis for your denial, to the Nuclear Regulatory
Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the
Regional Administrator, Region III, the Director, Office of Enforcement, United States Nuclear
Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the
Duane Arnold Energy Center.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a
response within 30 days of the date of this inspection report, with the basis for your
disagreement, to the Regional Administrator, Region III; and the NRC Resident Inspector at the
Duane Arnold Energy Center.
P. Wells
-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 (PARS) component
of NRC's Agencywide Document Access and Management System (ADAMS). ADAMS is
accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html
(the Public Electronic Reading Room).
Sincerely,
/RA/
Mark Ring, Branch Chief
Branch 1
Division of Reactor Projects
Docket No. 50-331
License No. DPR-49
Enclosure:
Inspection Report 05000331/2011005
w/Attachment: Supplemental Information
cc w/encl:
Distribution via ListServ
Enclosure
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No:
50-331
License No:
Report No:
Licensee:
NextEra Energy Duane Arnold, LLC
Facility:
Duane Arnold Energy Center
Location:
Palo, IA
Dates:
October 1 through December 31, 2011
Inspectors:
L. Haeg, Senior Resident Inspector
R. Murray, Resident Inspector
R. Orlikowski, Project Engineer
R. Edwards, Reactor Inspector
J. Beavers, Emergency Preparedness Inspector
M. Mitchell, Health Physicist
Approved by:
Mark Ring, Branch Chief
Branch 1
Division of Reactor Projects
Enclosure
TABLE OF CONTENTS
SUMMARY OF FINDINGS ........................................................................................................... 1
REPORT DETAILS ....................................................................................................................... 4
Summary of Plant Status ........................................................................................................... 4
1.
REACTOR SAFETY ....................................................................................................... 4
1R01
Adverse Weather Protection (71111.01) ............................................................. 4
1R04
Equipment Alignment (71111.04) ........................................................................ 5
1R05
Fire Protection (71111.05) .................................................................................. 5
1R11
Licensed Operator Requalification Program (71111.11) ..................................... 6
1R12
Maintenance Effectiveness (71111.12) ............................................................... 7
1R13
Maintenance Risk Assessments and Emergent Work Control (71111.13) ......... 8
1R15
Operability Determinations and Functionality Assessments (71111.15) ............. 9
1R19
Post-Maintenance Testing (71111.19) .............................................................. 10
1R22
Surveillance Testing (71111.22) ....................................................................... 11
1EP4
Emergency Action Level and Emergency Plan Changes (71114.04) ............... 12
1EP6
Drill Evaluation (71114.06) ................................................................................ 13
2.
RADIATION SAFETY ................................................................................................... 13
2RS1
Radiological Hazard Assessment and Exposure Controls (71124.01) ............. 13
2RS2
Occupational As-Low-As-Is-Reasonably-Achievable (ALARA) Planning and
Controls (71124.02) .......................................................................................... 19
4.
OTHER ACTIVITIES ..................................................................................................... 20
4OA1
Performance Indicator Verification (71151) ....................................................... 20
4OA2
Identification and Resolution of Problems (71152) ........................................... 22
4OA3
Follow-Up of Events and Notices of Enforcement Discretion (71153) .............. 27
4OA5
Other Activities .................................................................................................. 28
4OA6
Management Meetings ...................................................................................... 29
4OA7
Licensee-Identified Violations ........................................................................... 29
SUPPLEMENTAL INFORMATION ............................................................................................... 1
Key Points of Contact ................................................................................................................ 1
List of Items Opened, Closed and Discussed ............................................................................ 2
List of Documents Reviewed ..................................................................................................... 3
List of Acronyms Used .............................................................................................................. 9
1
Enclosure
SUMMARY OF FINDINGS
IR 05000331/2011005, 10/01/2011 - 12/31/2011; Duane Arnold Energy Center; Operability
Determinations and Functionality Assessments and Identification and Resolution of Problems.
This report covers a three-month period of inspection by resident inspectors and announced
baseline inspections by regional inspectors. One Severity Level IV violation was identified by
the inspectors and two Green findings were self-revealed. The violation and findings were
considered NCVs of NRC regulations. 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 and Violations
Cornerstone: Initiating Events
Green. A finding of very low safety significance and associated NCV of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed on
August 11, 2011, when both river water supply subsystems were rendered inoperable
following a sediment intrusion event. Specifically, the cause of the event was attributed
to the cancellation of a river bed survey that would have identified the increased
sediment buildup requiring increased monitoring and corrective actions (dredging, sand
pumping, and/or structural repairs). The cancellation of the river bed survey work order
was contrary to the requirements of Administrative Control Procedure 1208.3,
Preventive Maintenance Program, that required management approval prior to
cancelling the work order that was tied to the corrective action program. This issue of
concern was documented in the licensees corrective action program as condition report
01676836. Corrective actions included revision to affected river survey work orders to
ensure that they could not be cancelled without adequate review and approval, and
completion of river dredging and repairs to the upstream spur dikes.
The inspectors determined that the issue of concern represented a performance
deficiency because it was the result of the licensees failure to meet a procedural
requirement, and the cause was reasonably within the licensees ability to foresee and
correct and should have been prevented. The performance deficiency was determined
to be more than minor and a finding because it was associated with the Initiating Events
Cornerstone attribute of equipment performance, and it affected the cornerstone
objective to limit the likelihood of those events that upset plant stability and challenge
critical safety functions during power operations. The inspectors applied IMC 0609,
Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, to this
finding. Because the finding did not contribute to both the likelihood of a reactor trip and
the likelihood that mitigation equipment or functions would not be available under the
Initiating Events Cornerstone column of Table 4a, the finding was determined to be of
very low safety significance (Green). The inspectors determined that the contributing
cause that provided the most insight into the performance deficiency was associated
with the cross-cutting aspect of Human Performance, having Decision Making
components, and involving the licensee making safety or risk-significant decisions using
a systematic process, including formally defining the authority and roles for decisions
2
Enclosure
affecting nuclear safety. Specifically, several decisions were made with respect to spur
dike repairs and river monitoring; however, the requisite organizational reviews and
approvals associated with the river were not performed to ensure appropriate actions
were taken. H.1(a) (Section 4OA2.4)
Cornerstone: Barrier Integrity
Green. A finding of very low safety significance and associated NCV of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed on
October 31, 2011, when operators failed to follow Operating Instruction 999,
Reactor Building Crane. Specifically, this error resulted in the reactor building (RB)
crane striking the Independent Spent Fuel Storage Installation (ISFSI) inspection stand.
Immediate corrective actions included performing inspections of the dry storage
container transfer cask, ISFSI inspection stand, and reactor building crane.
The inspectors determined that attempting to move the crane over the ISFSI work
platform while the hand rails were installed was contrary to the RB crane operating
instruction and was an issue of concern. Failing to follow the RB crane operating
instruction was a performance deficiency because it was the result of the licensees
failure to meet a procedural requirement, and the cause was reasonably within the
licensees ability to foresee and correct and should have been prevented.
The performance deficiency was determined to be more than minor and a finding
because, if left uncorrected, the performance deficiency would have the potential to lead
to a more significant safety concern. Specifically, not following the RB crane operating
instructions could lead to a more significant event or cause damage to safety-related
equipment. The inspectors determined the finding could be evaluated using the SDP in
accordance with IMC 0609, Significance Determination Process, Attachment 0609.04,
Phase 1 - Initial Screening and Characterization of Findings, Table 4a for the Barrier
Integrity Cornerstone. Because the finding only affected the fuel barrier, the finding was
determined to be of very low safety significance (Green). The inspectors determined
that the contributing cause that provided the most insight into the performance deficiency
was associated with the cross-cutting aspect of Human Performance, having Work
Control components, and involving appropriately coordinating work activities by
incorporating actions to address the need to keep personnel apprised of work status, the
operational impact of work activities, and plant conditions that may affect work activities.
Specifically, the licensee did not implement appropriate work controls to ensure the hand
rails of the ISFSI inspection stand were removed prior to moving the crane for an activity
that was not associated with the ISFSI project. H.3(b) (Section 4OA2.5)
Cornerstone: Other
Severity Level IV. A Severity Level (SL) IV NCV of 10 CFR 50.72(b)(3)(v)(B) was
identified by the inspectors for the licensees failure to report within eight hours a
condition that, at the time of discovery, could have prevented the fulfillment of the
Residual Heat Removal (RHR) system Low Pressure Coolant Injection (LPCI) safety
function. Specifically, on December 2, 2011, a sizable void was identified in the B LPCI
discharge injection line resulting in the LPCI mode of RHR being declared inoperable.
The licensee documented the issue into their corrective action program (CAP), reported
the condition to the NRC on December 8, 2011, and, was in the process of reviewing the
cause of the issue to determine additional corrective actions.
3
Enclosure
The inspectors determined that the issue of concern represented a performance
deficiency because it was the result of the licensees failure to meet a regulatory
requirement, and the cause was reasonably within the licensees ability to foresee and
correct and should have been prevented. Because the performance deficiency is
considered to potentially impede or impact the ability of the NRC to perform its
regulatory oversight function, the performance deficiency was dispositioned using the
traditional enforcement process. Per NRC Enforcement Policy, Section 6.9.d.9, failing to
make a report required by 10 CFR 50.72 is categorized as an example of a Severity
Level IV violation. Additionally, because the violation was entered into the licensees
CAP, compliance was restored in a reasonable period of time, and was not repetitive or
willful; this violation is being treated as a non-cited SL IV violation, consistent with
Section 2.3.2 of the NRC Enforcement Policy. Because the performance deficiency was
not considered a finding using IMC 0612, Appendix B, Issue Screening, and did not
impact the Reactor Oversight Process Cornerstones of Safety, a cross-cutting aspect
was not assigned. (Section 1R15)
B.
Licensee-Identified Violations
A violation of very low safety significance that was identified by the licensee was
reviewed by inspectors. Corrective actions planned or taken by the licensee have been
entered into the licensees CAP. The violation and condition report is listed in
Section 4OA7 of this report.
4
Enclosure
REPORT DETAILS
Summary of Plant Status
Duane Arnold Energy Center (DAEC) operated at full power for the entire inspection period
except for brief down-power maneuvers to accomplish rod pattern adjustments and to conduct
planned surveillance testing activities.
1.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection (71111.01)
.1
Winter Seasonal Readiness Preparations
a.
Inspection Scope
The inspectors conducted a review of the licensees preparations for winter conditions to
verify that the plants design features and implementation of procedures were sufficient
to protect risk-significant systems from the effects of adverse weather. Documents for
the selected systems were reviewed to ensure that the systems would remain functional
when challenged by inclement weather. During the inspection, the inspectors focused
on plant-specific design features and the licensees procedures used to mitigate or
respond to adverse weather conditions. Additionally, the inspectors reviewed the
Updated Final Safety Analysis Report (UFSAR) and performance requirements for
systems selected for inspection, and verified that operator actions were appropriate as
specified by plant specific procedures. Cold weather protective components, such as
heat tracing and area heaters, were verified to be in operation where applicable.
The inspectors also reviewed CAP items to verify that the licensee was identifying
adverse weather issues at an appropriate threshold and entering them into the CAP in
accordance with station corrective action procedures. Specific documents reviewed
during this inspection are listed in the Attachment to this report. The inspectors reviews
focused specifically on the following plant systems due to their risk significance or
susceptibility to cold weather issues:
Control Building and Pump House Heating and Ventilation systems.
This inspection constituted one winter seasonal readiness preparations sample as
defined in Inspection Procedure (IP) 71111.01-05.
b.
Findings
No findings were identified.
5
Enclosure
1R04 Equipment Alignment (71111.04)
.1
Quarterly Partial System Walkdowns
a.
Inspection Scope
The inspectors performed partial system walkdowns of the following risk-significant
systems:
High Pressure Coolant Injection (HPCI) system; and
B Standby Diesel Generator (SBDG) and B Emergency Service Water (ESW)
subsystems during A SBDG surveillance testing.
The inspectors selected these systems based on their risk significance relative to the
Reactor Safety Cornerstones at the time they were inspected. The inspectors attempted
to identify any discrepancies that could impact the function of the system and, therefore,
potentially increase risk. The inspectors reviewed applicable operating procedures,
system diagrams, UFSAR, Technical Specification (TS) requirements, outstanding work
orders (WOs), condition reports (CRs), and the impact of ongoing work activities on
redundant trains of equipment in order to identify conditions that could have rendered
the systems incapable of performing their intended functions. The inspectors also
walked down accessible portions of the systems to verify system components and
support equipment were aligned correctly and operable. The inspectors examined the
material condition of the components and observed operating parameters of equipment
to verify that there were no obvious deficiencies. The inspectors also verified that the
licensee had properly identified and resolved equipment alignment problems that could
cause initiating events or impact the capability of mitigating systems or barriers and
entered them into the CAP with the appropriate significance characterization.
Documents reviewed are listed in the Attachment to this report.
This inspection constituted two quarterly partial system walkdown samples as defined in
b.
Findings
No findings were identified.
1R05 Fire Protection (71111.05)
.1
Routine Resident Inspector Tours (71111.05Q)
a.
Inspection Scope
The inspectors conducted fire protection walkdowns which were focused on availability,
accessibility, and the condition of firefighting equipment in the following risk-significant
plant areas:
Area Fire Plan (AFP) 03 and 07; Reactor Building - HPCI, Reactor Core Isolation
Cooling (RCIC); and Elevation 786 Corridor and Laydown Area;
AFP 74 and 79; Switchyard; and ISFSI;
6
Enclosure
AFP 4, 5 and 6; Reactor Building North Control Rod Drive (CRD) Module Area,
CRD repair and CRD Cable Rooms; Reactor Building South CRD Module Area
and Offgas Recombiner Rooms and Railroad Airlock; and Reactor Building RHR
Valve Room Elevation 7576;
AFP 10, 11 and 12; Main Exhaust Fan Room, Heating Hot Water Pump Room
and the Plant Air Supply Fan Room; Reactor Building Laydown Area Elevation
833-6; and Reactor Building Decay Tank and Condensate Phase Separator
Rooms; and
AFP 20; Aux Boiler Room, Emergency Diesel Generator Rooms and Generator
Day Tank Rooms Elevation 7576.
The inspectors reviewed areas to assess if the licensee had implemented a fire
protection program that adequately controlled combustibles and ignition sources within
the plant, effectively maintained fire detection and suppression capability, maintained
passive fire protection features in good material condition, and implemented adequate
compensatory measures for out-of-service, degraded or non-functional fire protection
equipment, systems, or features in accordance with the licensees fire plan.
The inspectors selected fire areas based on their overall contribution to internal fire risk
as documented in the plants Individual Plant Examination of External Events with later
additional insights, their potential to impact equipment which could initiate or mitigate a
plant transient, or their impact on the plants ability to respond to a security event.
Using the documents listed in the Attachment to this report, the inspectors verified that
fire hoses and extinguishers were in their designated locations and available for
immediate use; that fire detectors and sprinklers were unobstructed; that transient
material loading was within the analyzed limits; and fire doors, dampers, and penetration
seals appeared to be in satisfactory condition. The inspectors also verified that minor
issues identified during the inspection were entered into the licensees CAP.
Documents reviewed are listed in the Attachment to this report.
This inspection constituted five quarterly fire protection inspection samples as defined in
b.
Findings
No findings were identified.
1R11 Licensed Operator Requalification Program (71111.11)
.1
Resident Inspector Quarterly Review (71111.11Q)
a.
Inspection Scope
On October 27, 2011, the inspectors observed a crew of licensed operators in the plants
simulator during licensed operator requalification examinations to verify that operator
performance was adequate, evaluators were identifying and documenting crew
performance problems; and training was being conducted in accordance with licensee
procedures. The inspectors evaluated the following areas:
licensed operator performance;
crews clarity and formality of communications;
ability to take timely actions in the conservative direction;
7
Enclosure
prioritization, interpretation, and verification of annunciator alarms;
correct use and implementation of abnormal and emergency procedures;
control board manipulations;
oversight and direction from supervisors; and
ability to identify and implement appropriate TS actions and Emergency Plan
actions and notifications.
The crews performance in these areas was compared to pre-established operator action
expectations and successful critical task completion requirements. Documents reviewed
are listed in the Attachment to this report.
This inspection constituted one quarterly licensed operator requalification program
sample as defined in IP 71111.11.
b.
Findings
No findings were identified.
1R12 Maintenance Effectiveness (71111.12)
.1
Routine Quarterly Evaluations (71111.12Q)
a.
Inspection Scope
The inspectors evaluated the following:
DAEC Cycle 22 Periodic Evaluation; March 2, 2009, through December 9, 2010;
and
Feedwater Heater and Moisture Separator Drain Tank level control systems.
The inspectors reviewed events such as where ineffective equipment maintenance had
resulted in valid or invalid automatic actuations of engineered safeguards systems and
independently verified the licensee's actions to address system performance or condition
problems in terms of the following:
implementing appropriate work practices;
identifying and addressing common cause failures;
scoping of systems in accordance with 10 CFR 50.65(b) of the maintenance rule;
characterizing system reliability issues for performance;
charging unavailability for performance;
trending key parameters for condition monitoring;
ensuring 10 CFR 50.65(a)(1) or (a)(2) classification or re-classification; and
verifying appropriate performance criteria for structures, systems, and
components/functions classified as (a)(2), or appropriate and adequate goals and
corrective actions for systems classified as (a)(1).
The inspectors assessed performance issues with respect to the reliability, availability,
and condition monitoring of the system. In addition, the inspectors verified maintenance
effectiveness issues were entered into the CAP with the appropriate significance
characterization. Documents reviewed are listed in the Attachment to this report.
8
Enclosure
This inspection constituted two quarterly maintenance effectiveness samples as defined
in IP 71111.12-05.
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 the
maintenance and emergent work activities affecting risk-significant and safety-related
equipment listed below to verify that the appropriate risk assessments were performed
prior to removing equipment for work:
Work Week 1140 risk management during RHR system maintenance,
Electro-Hydraulic Control pump replacement, and ISFSI operations;
Work Week 1141 risk management during switchyard maintenance;
161 kV Tiffin Hills line out of service for transmission system operator
modifications;
Feedwater Heater and Moisture Separator Drain Tank operational decision
making issue;
RHR system LPCI function declared inoperable due to air voiding found in LPCI
injection piping; and
Inability to retract B Traversing In-core Probe (TIP) from reactor vessel.
These activities were selected based on their potential risk significance relative to the
Reactor Safety Cornerstones. As applicable for each activity, the inspectors verified that
risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate
and complete. When emergent work was performed, the inspectors verified that the
plant risk was promptly reassessed and managed. The inspectors reviewed the scope
of maintenance work, discussed the results of the assessment with the licensee's
probabilistic risk analyst or shift technical advisor, and verified plant conditions were
consistent with the risk assessment. The inspectors also reviewed TS requirements and
walked down portions of redundant safety systems, when applicable, to verify risk
analysis assumptions were valid and applicable requirements were met.
Documents reviewed are listed in the Attachment to this report.
This inspection constituted six maintenance risk assessments and emergent work
control samples as defined in IP 71111.13-05.
b.
Findings
No findings were identified.
9
Enclosure
1R15 Operability Determinations and Functionality Assessments (71111.15)
.1
Operability Evaluations
a.
Inspection Scope
The inspectors reviewed the following issues:
Operability evaluation for A SBDG system exhaust manifold leak;
Prompt Operability Determination (POD) for corroded Standby Gas Treatment
subsystem drain piping;
Intake structure ventilation system issues and impact on River Water Supply
(RWS) system;
River level instrumentation and intake structure sand gate position impact on
ultimate heat sink and RWS system operability; and
Void found in B RHR system LPCI discharge piping.
The inspectors selected these potential operability issues based on the risk significance
of the associated components and systems. The inspectors evaluated the technical
adequacy of the evaluations to ensure that TS operability was properly justified and the
subject component or system remained available such that no unrecognized increase in
risk occurred. The inspectors compared the operability and design criteria in the
appropriate sections of the TS and UFSAR with the licensees evaluations to determine
whether the components or systems were operable. Where compensatory measures
were required to maintain operability, the inspectors determined whether the measures
in place would function as intended and were properly controlled. The inspectors
determined, where appropriate, compliance with bounding limitations associated with the
evaluations. Additionally, the inspectors reviewed a sampling of CRs to verify that the
licensee was identifying and correcting any deficiencies associated with operability
evaluations. Documents reviewed are listed in the Attachment to this report.
This inspection constituted five operability determinations and functionality assessment
samples as defined in IP 71111.15-05.
b.
Findings
(1) Failure to Make Required Eight Hour Event Report per 10 CFR 50.72(b)(3)(v)(B)
Introduction: A Severity Level IV NCV of 10 CFR 50.72(b)(3)(v)(B) was identified by the
inspectors for the licensees failure to report within eight hours a condition that, at the
time of discovery, could have prevented the fulfillment of the RHR system LPCI safety
function.
Description: On November 18, 2011, the licensee performed a monthly surveillance test
to verify that the RHR system was full of water. The test consisted of, in part, static
venting of various portions of the system. During the venting, an abnormal amount of air
was vented from the B RHR LPCI injection piping. The licensee documented the
condition in the CAP and initiated a technical assessment for reportability (TAR) to
review past operability and reportability of the condition. At approximately 1300 hrs on
December 2, 2011, the licensee performed an ultrasonic examination of the B RHR
LPCI injection piping as part of the TAR evaluation and identified a 2-3 ft3 void within the
10
Enclosure
piping. After quantifying the void, the operations shift manager declared the LPCI
function of RHR inoperable and entered Technical Specification LCO 3.5.1, Condition B
for one low pressure emergency core cooling system subsystem inoperable (7 day
completion time to restore to an operable status). On December 4, 2011, while the
licensee continued to evaluate the significance of the void and determine corrective
actions, the inspectors were concerned that the void condition resulting in LPCI being
declared inoperable represented a condition that, at the time of discovery, could have
prevented the fulfillment of the RHR system LPCI safety function. Further, if it was a
condition that, at the time of discovery, could have prevented the fulfillment of the RHR
system LPCI safety function, the inspectors questioned why it wasnt reported to the
NRC within the eight-hour timeliness requirement of 10 CFR 50.72(b)(3)(v). The
licensee documented the inspectors questions and concerns in the CAP as CR
01714014, subsequently agreed that the condition was subject to an NRC report, and
made the report on December 8, 2011.
Analysis: The inspectors determined that the issue of concern represented a
performance deficiency because it was the result of the licensees failure to meet a
regulatory requirement, and the cause was reasonably within the licensees ability to
foresee and correct and should have been prevented. Because the performance
deficiency could potentially impede or impact the ability of the NRC to perform its
regulatory oversight function, the performance deficiency was dispositioned using the
traditional enforcement process. Per NRC Enforcement Policy, Section 6.9.d.9, failing to
make a report required by 10 CFR 50.72 is categorized as an example of a Severity
Level IV violation. Because the performance deficiency was not considered a finding
using IMC 0612, Appendix B, Issue Screening, and did not impact the Reactor
Oversight Process Cornerstones of Safety, a cross-cutting aspect was not assigned.
Enforcement: Title 10 CFR Part 50.72(b)(3)(v)(B), requires, in part, that operating
reactor licensees shall notify the NRC within eight hours of the occurrence of any event
or condition that at the time of discovery could have prevented the fulfillment of the
safety function of systems that are needed to remove residual heat. Contrary to this
requirement, on December 2, 2011, the licensee failed to report the void condition that at
the time of discovery could have prevented the fulfillment of the RHR system LPCI
safety function to the NRC within eight hours. Because the violation was entered into
the licensees CAP, compliance was restored in a reasonable period of time, and was
not repetitive or willful; this violation is being treated as a non-cited Severity Level IV
violation, consistent with Section 2.3.2 of the NRC Enforcement Policy.
(NCV 05000331/2011005-01, Failure to Make Required Eight Hour Event Report per
1R19 Post-Maintenance Testing (71111.19)
a.
Inspection Scope
The inspectors reviewed the following post-maintenance activities to verify that
procedures and test activities were adequate to ensure system operability and functional
capability:
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Enclosure
TIP primary containment isolation valve testing;
Standby and T1 transformers, and M breaker testing;
Main feedwater Leading Edge Flow Meter (LEFM) system testing; and
Intake structure ventilation system testing.
These activities were selected based upon the structure, system, or component's ability
to impact risk. The inspectors evaluated these activities for the following (as applicable):
the effect of testing on the plant had been adequately addressed; testing was adequate
for the maintenance performed; acceptance criteria were clear and demonstrated
operational readiness; test instrumentation was appropriate; tests were performed as
written in accordance with properly reviewed and approved procedures; equipment was
returned to its operational status following testing (temporary modifications or jumpers
required for test performance were properly removed after test completion); and test
documentation was properly evaluated. The inspectors evaluated the activities against
TSs, the UFSAR, 10 CFR Part 50 requirements, licensee procedures, and various
NRC generic communications to ensure that the test results adequately ensured that the
equipment met the licensing basis and design requirements. In addition, the inspectors
reviewed corrective action documents associated with post-maintenance tests to
determine whether the licensee was identifying problems and entering them in the CAP
and that the problems were being corrected commensurate with their importance to
safety. Documents reviewed are listed in the Attachment to this report.
This inspection constituted four post-maintenance testing samples as defined in
b.
Findings
No findings were identified.
1R22 Surveillance Testing (71111.22)
a.
Inspection Scope
The inspectors reviewed the test results for the following activities to determine whether
risk-significant systems and equipment were capable of performing their intended safety
function and to verify testing was conducted in accordance with applicable procedural
and TS requirements:
Surveillance Test Procedure (STP) 3.8.1-04B; B Standby Diesel Generator
Operability Test (Slow Start from Normal Starting Air) (Routine);
STP 3.6.1.6-01; Pressure Suppression Chamber to Reactor Building Vacuum
Breaker Operability Test (In-service test);
STP 3.7.4-01B; B Standby Filter Unit - Logic System Functional Test and
Simulated Automatic Actuation (Routine);
STP NS300002; Tracer Gas Test of Control Building Envelope (Routine); and
STP 3.5.3-04; RCIC Simulated Auto Actuation Test (Routine).
The inspectors observed in-plant activities and reviewed procedures and associated
records to determine the following:
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Enclosure
did preconditioning occur;
were the effects of the testing adequately addressed by control room personnel
or engineers prior to the commencement of the testing;
were acceptance criteria clearly stated, demonstrated operational readiness, and
consistent with the system design basis;
plant equipment calibration was correct, accurate, and properly documented;
as-left setpoints were within required ranges; and the calibration frequency was
in accordance with TSs, the UFSAR, procedures, and applicable commitments;
measuring and test equipment calibration was current;
test equipment was used within the required range and accuracy; applicable
prerequisites described in the test procedures were satisfied;
test frequencies met TS requirements to demonstrate operability and reliability;
tests were performed in accordance with the test procedures and other
applicable procedures; jumpers and lifted leads were controlled and restored
where used;
test data and results were accurate, complete, within limits, and valid;
test equipment was removed after testing;
where applicable for inservice testing activities, testing was performed in
accordance with the applicable version of Section XI, American Society of
Mechanical Engineers Code, and reference values were consistent with the
system design basis;
where applicable, test results not meeting acceptance criteria were addressed
with an adequate operability evaluation or the system or component was
declared inoperable;
where applicable for safety-related instrument control surveillance tests,
reference setting data were accurately incorporated in the test procedure;
where applicable, actual conditions encountering high resistance electrical
contacts were such that the intended safety function could still be accomplished;
prior procedure changes had not provided an opportunity to identify problems
encountered during the performance of the surveillance or calibration test;
equipment was returned to a position or status required to support the
performance of its safety functions; and
all problems identified during the testing were appropriately documented and
dispositioned in the CAP.
Documents reviewed are listed in the Attachment to this report.
This inspection constituted four routine surveillance testing samples, and one in-service
test sample as defined in IP 71111.22, Sections -02 and -05.
b.
Findings
No findings were identified.
1EP4 Emergency Action Level and Emergency Plan Changes (71114.04)
a.
Inspection Scope
Since the last NRC inspection of this program area, revisions of the Emergency Plan
and of the Emergency Action Levels were implemented based on the licensees
13
Enclosure
determination, in accordance with 10 CFR 50.54(q), that the changes resulted in no
decrease in effectiveness of the Plan and that the revised Plan as changed continued to
meet the requirements of 10 CFR 50.47(b) and Appendix E to 10 CFR Part 50. The
inspectors conducted a sampling review of the Emergency Plan changes and a review of
the Emergency Action Level changes made between December 2010 and September
2011 to evaluate for potential decreases in effectiveness of the Plan. However, this
review does not constitute formal NRC approval of the changes. Therefore, these
changes remain subject to future NRC inspection in their entirety.
This inspection constituted one emergency action level and emergency plan changes
inspection sample as defined in IP 71114.04-05.
b.
Findings
No findings were identified.
1EP6 Drill Evaluation (71114.06)
.1
Emergency Preparedness Drill Observation
a.
Inspection Scope
The inspectors evaluated the conduct of a routine licensee emergency drill on
November 9, 2011, to identify any weaknesses and deficiencies in classification,
notification, and protective action recommendation development activities.
The inspectors observed emergency response operations in the Control Room
Simulator and the Technical Support Center to determine whether the event
classification, notifications, and protective action recommendations were performed in
accordance with procedures. The inspectors also attended the licensee drill critique to
compare any inspector-observed weakness with those identified by the licensee staff
in order to evaluate the critique and to verify whether the licensee staff was properly
identifying weaknesses and entering them into the corrective action program. As part
of the inspection, the inspectors reviewed the drill package and other documents listed
in the Attachment to this report.
This inspection constituted one emergency preparedness drill inspection sample as
defined in IP 71114.06-05.
b.
Findings
No findings were identified.
2.
RADIATION SAFETY
Cornerstones: Occupational and Public Radiation Safety
2RS1 Radiological Hazard Assessment and Exposure Controls (71124.01)
This inspection constituted one complete radiological hazard assessment and exposure
controls sample as defined in IP 71124.01-05.
14
Enclosure
.1
Inspection Planning (02.01)
a.
Inspection Scope
The inspectors reviewed all licensee performance indicators for the occupational
exposure cornerstone for follow-up. The inspectors reviewed the results of radiation
protection program audits (e.g., licensees quality assurance audits or other independent
audits). The inspectors reviewed any reports of operational occurrences related to
occupational radiation safety since the last inspection. The inspectors reviewed the
results of the audit and operational report reviews to gain insights into overall licensee
performance.
b.
Findings
No findings were identified.
.2
Radiological Hazard Assessment (02.02)
a.
Inspection Scope
The inspectors determined if there had been any changes to plant operations since the
last inspection that could have resulted in a significant new radiological hazard for onsite
workers or members of the public. The inspectors evaluated whether the licensee
assessed the potential impact of any changes and had implemented periodic monitoring,
as appropriate, to detect and quantify the radiological hazard(s).
The inspectors reviewed the last two radiological surveys from selected plant areas and
evaluated whether the thoroughness and frequency of the surveys where appropriate for
the given radiological hazard(s).
The inspectors conducted walkdowns of the facility, including radioactive waste
processing, storage, and handling areas to evaluate material conditions and performed
independent radiation measurements to verify conditions.
The inspectors selected the following radiologically risk-significant work activities that
involved exposure to radiation:
Entry into Heater/Condenser Bay.
For these work activities, the inspectors assessed whether the pre-work surveys
performed were appropriate to identify and quantify the radiological hazard(s) and to
establish adequate protective measures. The inspectors evaluated the radiological
survey program to determine if hazards were properly identified, including the following:
identification of hot particles;
the presence of alpha emitters;
the potential for airborne radioactive materials, including the potential presence
of transuranics and/or other hard-to-detect radioactive materials;
the hazards associated with work activities that could suddenly and severely
increase radiological conditions and that the licensee had established a means to
inform workers of changes that could have significantly impacted their
occupational dose; and
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Enclosure
severe radiation field dose gradients that could have resulted in non-uniform
exposures of the body.
The inspectors observed work in potential airborne areas and evaluated whether the air
samples were representative of the breathing air zone. The inspectors evaluated
whether continuous air monitors were located in areas with low background to minimize
false alarms and were representative of actual work areas. The inspectors evaluated
the licensees program for monitoring levels of loose surface contamination in areas of
the plant with the potential for the contamination to become airborne.
b.
Findings
No findings were identified.
.3
Instructions to Workers (02.03)
a.
Inspection Scope
The inspectors selected various containers holding non-exempt licensed radioactive
materials that could have caused unplanned or inadvertent exposure of workers, and
assessed whether the containers were labeled and controlled in accordance with
10 CFR 20.1904, Labeling Containers, or met the requirements of 10 CFR 20.1905(g),
Exemptions To Labeling Requirements.
The inspectors reviewed the following radiation work permits used to access high
radiation areas and evaluated the specified work control instructions or control barriers.
In-Service Project - N2 Penetration Work;
Segregation of Hot Trash and Vacuum Cleaner High Efficiency Particulate Air
Filter Change Out; and
Traversing In-core Probe Room Maintenance.
For these radiation work permits, the inspectors assessed whether allowable stay times
or permissible dose (including from the intake of radioactive material) for radiologically
significant work under each radiation work permit were clearly identified. The inspectors
evaluated whether electronic personal dosimeter alarm set-points were in conformance
with survey indications and plant policy.
The inspectors reviewed selected occurrences where a workers electronic personal
dosimeter noticeably malfunctioned or alarmed. The inspectors evaluated whether
workers had responded appropriately to the off-normal condition. The inspectors
assessed whether the issue was included in the corrective action program and dose
evaluations were conducted as appropriate.
For work activities that could suddenly and severely increase radiological conditions,
the inspectors assessed the licensees means to inform workers of changes that could
significantly impact their occupational dose.
b.
Findings
No findings were identified.
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Enclosure
.4
Contamination and Radioactive Material Control (02.04)
a.
Inspection Scope
The inspectors observed locations where the licensee monitored potentially
contaminated material leaving the radiological control area and inspected the methods
used for control, survey, and release from these areas. The inspectors observed the
performance of personnel surveying and releasing material for unrestricted use and
evaluated whether the work was performed in accordance with plant procedures and
whether the procedures were sufficient to control the spread of contamination and
prevent unintended release of radioactive materials from the site. The inspectors
assessed whether the radiation monitoring instrumentation had appropriate sensitivity for
the types of radiation present.
The inspectors reviewed the licensees criteria for the survey and release of potentially
contaminated material. The inspectors evaluated whether there was guidance on how to
respond to alarms that would indicate the presence of licensed radioactive material.
The inspectors reviewed the licensees procedures and records to verify that the
radiation detection instrumentation was used at its typical sensitivity level based on
appropriate counting parameters. The inspectors assessed whether or not the licensee
has established a de facto release limit by altering the instruments typical sensitivity
through such methods as raising the energy discriminator level or locating the instrument
in a high-radiation background area.
The inspectors selected several sealed sources from the licensees inventory records
and assessed whether the sources were accounted for and verified to be intact.
The inspectors evaluated whether any transactions, since the last inspection, involving
nationally tracked sources were reported in accordance with 10 CFR 20.2207.
b.
Findings
No findings were identified.
.5
Radiological Hazards Control and Work Coverage (02.05)
a.
Inspection Scope
The inspectors evaluated ambient radiological conditions (e.g., radiation levels or
potential radiation levels) during tours of the facility. The inspectors assessed whether
the conditions were consistent with applicable posted surveys, radiation work permits,
and worker briefings.
The inspectors evaluated the adequacy of radiological controls, such as required
surveys, radiation protection job coverage (including audio and visual surveillance for
remote job coverage), and contamination controls. The inspectors evaluated the
licensees use of electronic personal dosimeters in high noise areas as high radiation
area monitoring devices.
The inspectors assessed whether radiation monitoring devices were placed on the
individuals body consistent with licensee procedures. The inspectors assessed whether
17
Enclosure
the dosimeters were placed in the location of highest expected dose or that the licensee
properly employed an NRC-approved method of determining effective dose equivalent.
The inspectors reviewed the application of dosimetry to effectively monitor exposure to
personnel in high-radiation work areas with significant dose rate gradients.
The inspectors reviewed the following radiation work permits for work within airborne
radioactivity areas with the potential for individual worker internal exposures.
Segregation of Hot Trash and Vacuum Cleaner High Efficiency Particulate Air
Filter Change Out;
Dry Fuel Storage Project; and
Cask Pit Clean-up and Transport of Tri-Nuc 260 Hoses and Filters.
For these radiation work permits, the inspectors evaluated airborne radioactive
controls and monitoring, including potential for significant airborne levels
(e.g., grinding, grit blasting, system breaches, entry into tanks, cubicles, and
reactor cavities). The inspectors assessed barrier (e.g., tent or glove box) integrity and
temporary high-efficiency particulate air ventilation system operation. The inspectors
examined the licensees physical and programmatic controls for highly activated or
contaminated materials (nonfuel) stored within spent fuel and other storage pools.
The inspectors assessed whether appropriate controls (i.e., administrative and physical
controls) were in place to preclude inadvertent removal of these materials from the pool.
The inspectors examined the posting and physical controls for selected high radiation
areas and very high radiation areas to verify conformance with the occupational
performance indicator.
b.
Findings
No findings were identified.
.6
Risk-Significant High Radiation Area and Very High Radiation Area Controls (02.06)
a.
Inspection Scope
The inspectors discussed with the radiation protection manager the controls and
procedures for high-risk high radiation areas and very high radiation areas.
The inspectors discussed methods employed by the licensee to provide stricter control
of very high radiation area access as specified in 10 CFR 20.1602, Control of Access to
Very High Radiation Areas, and Regulatory Guide 8.38, Control of Access to High and
Very High Radiation Areas of Nuclear Plants. The inspectors assessed whether any
changes to licensee procedures substantially reduced the effectiveness and/or level of
worker protection.
The inspectors discussed the controls in place for special areas that have the potential
to become very high radiation areas during certain plant operations with first-line health
physics supervisors (or equivalent positions having backshift health physics oversight
authority). The inspectors assessed whether these plant operations required
communication beforehand with the health physics group, so as to allow corresponding
timely actions to properly post, control, and monitor the radiation hazards including
re-access authorization.
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Enclosure
The inspectors evaluated licensee controls for very high radiation areas and areas with
the potential to become very high radiation areas to ensure that individuals were not able
to gain unauthorized access to the very high radiation area.
b.
Findings
No findings were identified.
.7
Radiation Worker Performance (02.07)
a.
Inspection Scope
The inspectors observed radiation worker performance with respect to stated radiation
protection work requirements. The inspectors assessed whether workers were aware of
the radiological conditions in their workplace and the radiation work permit controls/limits
in place, and whether their performance reflected the level of radiological hazards
present.
The inspectors reviewed radiological problem reports since the last inspection that found
the cause of the event to be human performance errors. The inspectors evaluated
whether there was an observable pattern traceable to a similar cause. The inspectors
assessed whether this perspective matched the corrective action approach taken by the
licensee to resolve the reported problems. The inspectors discussed with the radiation
protection manager any problems with the corrective actions planned or taken.
b.
Findings
No findings were identified.
.8
Radiation Protection Technician Proficiency (02.08)
a.
Inspection Scope
The inspectors observed the performance of the radiation protection technicians with
respect to all radiation protection work requirements. The inspectors evaluated whether
technicians were aware of the radiological conditions in their workplace and the radiation
work permit controls/limits, and whether their performance was consistent with their
training and qualifications with respect to the radiological hazards and work activities.
The inspectors reviewed radiological problem reports since the last inspection that found
the cause of the event to be radiation protection technician error. The inspectors
evaluated whether there was an observable pattern traceable to a similar cause.
The inspectors assessed whether this perspective matched the corrective action
approach taken by the licensee to resolve the reported problems.
b.
Findings
No findings were identified.
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Enclosure
.9
Problem Identification and Resolution (02.09)
a.
Inspection Scope
The inspectors evaluated whether problems associated with radiation monitoring and
exposure control 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 that involve radiation monitoring and
exposure controls. The inspectors assessed the licensees process for applying
operating experience to their plant.
b.
Findings
No findings were identified.
2RS2 Occupational As-Low-As-Is-Reasonably-Achievable (ALARA) Planning and Controls
(71124.02)
These inspection activities supplement those documented in Inspection Report 05000331/2010004, and constitute one complete occupational ALARA planning and
controls sample as defined in IP 71124.01-05.
.1
Inspection Planning (02.01)
a.
Inspection Scope
The inspectors reviewed pertinent information regarding plant collective exposure
history, current exposure trends, and ongoing or planned activities in order to assess
current performance and exposure challenges. The inspectors reviewed the plants
three year rolling average collective exposure.
The inspectors reviewed the site-specific trends in collective exposures
(using NUREG-0713, Occupational Radiation Exposure at Commercial Nuclear Power
Reactors and Other Facilities, and plant historical data) and source term (average
contact dose rate with reactor coolant piping) measurements (using Electric Power
Research Institute) TR-108737, BWR Iron Control Monitoring Interim Report, issued
December 1998, and/or plant historical data, when available).
The inspectors reviewed site-specific procedures associated with maintaining
occupational exposures ALARA, which included a review of processes used to estimate
and track exposures from specific work activities.
b.
Findings
No findings were identified.
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Enclosure
.2
Radiological Work Planning (02.02)
a.
Inspection Scope
The inspectors compared the results achieved (dose rate reductions, person-rem used)
with the intended dose established in the licensees ALARA planning for these work
activities. The inspectors compared the person-hour estimates provided by maintenance
planning and other groups to the radiation protection group with the actual work activity
time requirements, and evaluated the accuracy of these time estimates. The inspectors
assessed the reasons (e.g., failure to adequately plan the activity, failure to provide
sufficient work controls) for any inconsistencies between intended and actual work
activity doses.
The inspectors determined whether post-job reviews were conducted and if identified
problems were entered into the licensees corrective action program.
b.
Findings
No findings were identified.
4.
OTHER ACTIVITIES
Cornerstones: Mitigating Systems and Barrier Integrity
4OA1 Performance Indicator Verification (71151)
.1
Mitigating Systems Performance Index - Residual Heat Removal System
a.
Inspection Scope
The inspectors sampled licensee submittals for the Mitigating Systems Performance
Index (MSPI) - Residual Heat Removal System Performance Indicator (PI) for the period
from the fourth quarter 2010 through the third quarter 2011. To determine the accuracy
of the PI data reported during those periods, PI definitions and guidance contained in
Nuclear Energy Institute (NEI) Document 99-02, Regulatory Assessment Performance
Indicator Guideline, Revision 6, was used. The inspectors reviewed the licensees
operator narrative logs, condition reports, MSPI derivation reports, event reports and
NRC Integrated Inspection Reports for the period of October 2010 through September
2011 to validate the accuracy of the submittals. The inspectors reviewed the MSPI
component risk coefficient to determine if it had changed by more than 25 percent 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 CAP to determine
if any problems had been identified with the PI data collected or transmitted for this
indicator and none were identified. Documents reviewed are listed in the Attachment to
this report.
This inspection constituted one MSPI residual heat removal system sample as defined in
b.
Findings
No findings were identified.
21
Enclosure
.2
Mitigating Systems Performance Index - Cooling Water Systems
a.
Inspection Scope
The inspectors sampled licensee submittals for the Mitigating Systems Performance
Index - Cooling Water Systems PI for the period from the fourth quarter 2010 through the
third quarter 2011. To determine the accuracy of the PI data reported during those
periods, PI definitions and guidance contained in NEI Document 99-02, Regulatory
Assessment Performance Indicator Guideline, Revision 6, was used. The inspectors
reviewed the licensees operator narrative logs, condition reports, MSPI derivation
reports, event reports and NRC Integrated Inspection Reports for the period of October
2010 through September 2011 to validate the accuracy of the submittals. The
inspectors reviewed the MSPI component risk coefficient to determine if it had changed
by more than 25 percent 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 CAP to determine if any problems had been identified with the PI data
collected or transmitted for this indicator and none were identified. Documents reviewed
are listed in the Attachment to this report.
This inspection constituted one MSPI cooling water system sample as defined in
b.
Findings
No findings were identified.
.3
Safety System Functional Failures
a.
Inspection Scope
The inspectors sampled licensee submittals for the Safety System Functional Failures PI
for the period from the fourth quarter 2010 through the third quarter 2011. To determine
the accuracy of the PI data reported during those periods, PI definitions and guidance
contained in NEI Document 99-02, Regulatory Assessment Performance Indicator
Guideline, Revision 6, and NUREG-1022, Event Reporting Guidelines 10 CFR 50.72
and 50.73" definitions and guidance, were used. The inspectors reviewed the licensees
operator narrative logs, operability assessments, maintenance rule records,
maintenance work orders, condition reports, event reports and NRC Integrated
Inspection Reports for the period of October 2010 through September 2011 to validate
the accuracy of the submittals. The inspectors also reviewed the licensees CAP to
determine if any problems had been identified with the PI data collected or transmitted
for this indicator and none were identified. Documents reviewed are listed in the
Attachment to this report.
This inspection constituted one safety system functional failures PI sample as defined in
b.
Findings
No findings were identified.
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Enclosure
4OA2 Identification and Resolution of Problems (71152)
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency
Preparedness, Public Radiation Safety, Occupational Radiation Safety, and
.1
Routine Review of Items Entered into the Corrective Action Program
a.
Inspection Scope
As part of the various baseline inspection procedures 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 CAP at
an appropriate threshold, that adequate attention was being given to timely corrective
actions, and that adverse trends were identified and addressed. Attributes reviewed
included: identification of the problem was complete and accurate; timeliness was
commensurate with the safety significance; evaluation and disposition of performance
issues, generic implications, common causes, contributing factors, root causes,
extent-of-condition reviews, and previous occurrences reviews were proper and
adequate; and that the classification, prioritization, focus, and timeliness of corrective
actions were commensurate with safety and sufficient to prevent recurrence of the issue.
Minor issues entered into the licensees CAP as a result of the inspectors observations
are included in the Attachment to this report.
These routine reviews for the identification and resolution of problems did not constitute
any additional inspection samples. Instead, by procedure, they were considered an
integral part of the inspections performed during the quarter and documented in
Section 1 of this report.
b.
Findings
No findings were identified.
.2
Daily Corrective Action Program Reviews
a.
Inspection Scope
In order to assist with the identification of repetitive equipment failures and specific
human performance issues for follow-up, the inspectors performed a daily screening of
items entered into the licensees CAP. This review was accomplished through
inspection of the stations daily condition report packages.
These daily reviews were performed by procedure as part of the inspectors daily plant
status monitoring activities and, as such, did not constitute any separate inspection
samples.
b.
Findings
No findings were identified.
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Enclosure
.3
Semi-Annual Trend Review
a.
Inspection Scope
The inspectors performed a review of the licensees CAP and associated documents to
identify trends that could indicate the existence of a more significant safety issue.
The inspectors review was focused on repetitive equipment issues, but also considered
the results of daily inspector CAP item screening discussed in Section 4OA2.2 above,
licensee trending efforts, and licensee human performance results. The inspectors
review nominally considered the six month period of June 2011 through November 2011,
although some examples expanded beyond those dates where the scope of the trend
warranted.
The review also included issues documented outside the normal CAP in major
equipment problem lists, repetitive and/or rework maintenance lists, departmental
problem/challenges lists, system health reports, quality assurance audit/surveillance
reports, self-assessment reports, and Maintenance Rule assessments. The inspectors
compared and contrasted their results with the results contained in the licensees
CAP trending reports. Corrective actions associated with a sample of the issues
identified in the licensees trending reports were reviewed for adequacy.
This inspection constituted one semi-annual trend review sample as defined in
b.
Findings
No findings were identified.
.4
Selected Issue Follow-Up Inspection: Root Cause Evaluation 01676836, Both River
Water Supply (RWS) Subsystems Inoperable
a.
Inspection Scope
As a follow-up to the unplanned shutdown of the plant on August 11, 2011, due to the
RWS system being declared inoperable, the inspectors reviewed the root cause
evaluation (RCE) performed by the licensee. This review was to determine whether the
causal factors, contributing factors, and corrective actions were appropriate for the
circumstances surrounding the event. Based on the inspectors review of the RCE
several issues of concern were noted, one of which represented a self-revealed
performance deficiency discussed below. Overall, the inspectors concluded that the root
cause evaluation was performed in a thorough, probing manner; and several corrective
actions were identified that should be appropriate to prevent recurrence of the event.
This inspection constituted one selected issue follow-up sample as defined in
24
Enclosure
b.
Findings
(1) Cancellation of River Survey Work Order Causes Inoperability of River Water Supply
System
Introduction: A finding of very low safety significance and associated NCV of
10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings,
was self-revealed on August 11, 2011, when both river water supply subsystems were
rendered inoperable following a sediment intrusion event.
Description: On August 11, 2011, the licensee entered a forced outage after declaring
the RWS system inoperable due to a sand intrusion event that rendered both intake
structure travelling screens non-functional (see NRC Inspection Report 05000331/2011004 for additional information). The licensee performed a RCE for the
event and determined several causal factors were involved. The inspectors reviewed
the RCE, specifically focusing on the two root causes that were determined by the
licensee. Root Cause #1 involved the decision to cancel a July 2011 river survey that
would have detected the degrading conditions of the ultimate heat sink. Root Cause #2
involved the failure to establish preventive maintenance in the form of a river survey
following high river flow or level conditions.
The inspectors focused their review of the licensees evaluation of Root Cause #1 and
its contributing causes since it represented a self-revealed issue of concern and should
have been prevented. The evaluation determined that the July 2011 river survey would
have identified the increased sediment buildup upstream of the intake due to
degradation of upstream spur dikes. This survey would have triggered increased
monitoring and corrective actions in the form of dredging, sand pumping, and/or
structural repairs to the spur dikes themselves. The RCE also determined that the
July 2011 river survey WO 40056778 was cancelled without receiving review and
approval per the requirements of Administrative Control Procedure (ACP) 1208.3,
Preventive Maintenance Program. This procedure specifically required that
management approval was required prior to cancelling any work order that was tied to
the corrective action program. In the case of WO 40056778, it was a corrective action to
prevent recurrence from a RCE performed in 2003.
Analysis: The inspectors determined that the issue of concern represented a
performance deficiency because it was the result of the licensees failure to meet a
procedural requirement, and the cause was reasonably within the licensees ability to
foresee and correct and should have been prevented. The performance deficiency was
determined to be more than minor and a finding because it was associated with the
Initiating Events Cornerstone attribute of equipment performance, and it affected the
cornerstone objective to limit the likelihood of those events that upset plant stability and
challenge critical safety functions during power operations. The inspectors applied
IMC 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings,
to this finding. Because the finding did not contribute to both the likelihood of a reactor
trip and the likelihood that mitigation equipment or functions would not be available
under the Initiating Events Cornerstone column of Table 4a, the finding was determined
to be of very low safety significance (Green).
The inspectors determined that the contributing cause that provided the most insight
into the performance deficiency was associated with the cross-cutting aspect of
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Enclosure
Human Performance, having Decision Making components, and involving the licensee
making safety or risk-significant decisions using a systematic process, including formally
defining the authority and roles for decisions affecting nuclear safety. Specifically,
risk-significant decisions were made with respect to cancelling a river survey and
deferring repairs to the spur dikes; however, the requisite organizational reviews and
approvals associated with the ultimate heat sink were not performed to ensure
appropriate actions were taken. H.1(a)
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures,
and Drawings, requires, in part, that activities affecting quality shall be prescribed by
documented procedures of a type appropriate to the circumstances and shall be
accomplished in accordance with these procedures. Procedure ACP 1208.3,
Preventive Maintenance Program, establishes, in part, the licensees implementing
procedure for approving and cancelling work at the facility.
Contrary to the above, on July 19, 2011, the licensee failed to accomplish ACP 1208.3,
Section 3.11, which required, in part, Management Review Committee review of
WO 40056778 prior to voiding the WO for any reason. Corrective actions included
revision to affected river survey work orders to ensure that they could not be cancelled
without adequate review and approval, and completion of river dredging and repairs to
the upstream spur dikes. Because this violation was of very low safety significance and
was entered into the licensees corrective action program as CR 01676836, the violation
is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement
Policy (NCV 05000331/2011005-02, Cancellation of River Survey Work Order
Causes Inoperability of River Water Supply System).
.5
Selected Issue Follow-Up Inspection: Reactor Building Crane Strikes ISFSI Inspection
Stand
a.
Inspection Scope
During a review of items entered in the licensees CAP, the inspectors recognized a
condition report documenting displacement and potential damage to the ISFSI
inspection stand and reactor building (RB) crane control cab. The inspectors followed
the licensees immediate corrective actions, including their follow up inspections, since a
loaded ISFSI transfer cask was in the process of vacuum drying. The inspectors also
followed the licensees assessment of damage to the RB crane structural integrity prior
to the licensee releasing the RB crane for use. The inspectors also reviewed a RCE
conducted by the licensee for the event.
This inspection constituted one selected issue follow-up sample as defined in
b.
Findings
(1) Procedural Non-Compliance Results in RB Crane Striking the ISFSI Inspection Stand
Introduction: A finding of very low safety significance and associated NCV of
10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was
self-revealed on October 31, 2011, when operators failed to follow Operating Instruction
(OI) 999, Reactor Building Crane. Specifically, this error resulted in the RB crane
striking the ISFSI inspection stand.
26
Enclosure
Description: On October 31, 2011, the licensee was in the midst of a several month long
ISFSI loading campaign. Following completion of the main work activities for the day, a
decision was made to move a storage cabinet from the first floor of the reactor building
to the refuel floor, where most of the ISFSI project activities were taking place. Prior to
moving the cabinet, operators and a contract supervisor held an informal pre-job brief to
discuss the lift. The work activity being performed was not governed by a work order to
perform the task. Later investigation revealed that due to the equipment used for the
evolution and the area the work was taking place, the work activity could be considered
minor maintenance and require a minor maintenance work request and order, at a
minimum. Following the informal brief, a crane operator went to the refuel floor to get
ready to lift the storage cabinet. In order to make the lift, the operator needed to move
the crane from the north end of the refuel floor to the south end. While traversing the
crane from north to south, the operator noticed that the hand rails for the ISFSI
inspection stand were installed. Unfortunately, the crane cab was only a couple feet
from the platform when the operator noticed this, and the operator was not able to stop
the crane before it struck the ISFSI inspection stand hand rails. The ISFSI inspection
stand was displaced several inches and was stopped by the dry storage canister (DSC)
transfer cask, which was recently loaded with spent fuel undergoing a vacuum drying
process. Subsequently, the licensee verified the drying process was uninterrupted and
the DSC transfer cask was not damaged. They also performed inspections of the
inspection stand and made repairs necessary to ensure personnel safety. Inspections of
structural integrity were also performed for the RB crane prior to additional crane
operations.
The licensee performed a RCE for the event and noted that the inspection stand hand
rail interference was first identified in 2003. Corrective actions for the interference
included adding a caution statement to OI 999, Reactor Building Crane, and
performance of a periodic inspection of the RB crane. In addition, the licensee placed
an information tag on the control panel of the crane to warn operators of the potential
inspection stand hand rail interference. This information tag was removed at one point
and not in place at the RB crane control panel prior to the event.
The inspectors questioned why the RCE did not specifically identify the apparently
ineffective corrective actions from 2003 as either a root or contributing cause.
Although the RCE discussed the ineffective corrective actions from 2003, the inspectors
concluded that the inadequate controls in place to maintain the information tag was a
contributing cause to the event. Also, the inspectors noted that the placement of the
information tag in the crane cab was not identified as a long-term corrective action and
following the ISFSI campaign in 2003, the tag was removed. There was no requirement
added to have the tag reinstalled when the ISFSI inspection stand was erected for the
2011 ISFSI campaign.
Analysis: The inspectors determined that attempting to move the crane over the ISFSI
work platform while the hand rails were installed was contrary to OI 999 and was an
issue of concern. Failing to follow OI 999 was a performance deficiency because it was
the result of the licensees failure to meet a procedural requirement, and the cause was
reasonably within the licensees ability to foresee and correct and should have been
prevented. The performance deficiency was determined to be more than minor and a
finding because, if left uncorrected, the performance deficiency would have the potential
to lead to a more significant safety concern. Specifically, not following OI 999 could lead
27
Enclosure
to more significant event or cause damage to safety-related equipment The inspectors
concluded this finding was associated with the Barrier Integrity Cornerstone.
The inspectors determined the finding could be evaluated using the SDP in accordance
with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 -
Initial Screening and Characterization of Findings, Table 4a for the Barrier Integrity
Cornerstone. Because the finding only affected the fuel barrier, the finding was
determined to be of very low safety significance (Green).
The inspectors determined that the contributing cause that provided the most insight into
the performance deficiency was associated with the cross-cutting aspect of Human
Performance, having Work Control components, and involving appropriately coordinating
work activities by incorporating actions to address the need to keep personnel apprised
of work status, the operational impact of work activities, and plant conditions that may
affect work activities. Specifically, the licensee did not implement appropriate work
controls to ensure the hand rails of the ISFSI inspection stand were removed prior to
moving the crane for an activity that was not associated with the ISFSI project. H.3(b)
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures,
and Drawings, requires, in part, that activities affecting quality shall be prescribed by
documented procedures of a type appropriate to the circumstances and shall be
accomplished in accordance with these procedures. Contrary to this, on
October 31, 2011, an operator failed to accomplish an activity affecting quality in
accordance with procedures. Specifically, the operator failed to comply with procedure
OI 999, Revision 39, Reactor Building Crane, which contained a caution statement
indicating that the crane cab would not clear the handrails of the ISFSI work platform
when the cask work platform is installed. Immediate corrective actions included
performing inspections of the DSC transfer cask, vacuum drying operations,
ISFSI inspection stand, and RB crane. Because this violation was of very low safety
significance and it was entered into the licensees corrective action program as
CR 1701934, this violation is being treated as an NCV, consistent with Section 2.3.2 of
the NRC Enforcement Policy (NCV 05000331/2011005-03, Procedural
Non-Compliance Results in RB Crane Colliding with ISFSI Inspection Stand).
4OA3 Follow-Up of Events and Notices of Enforcement Discretion (71153)
.1
Reactor Feedwater Pump Area Deluge Inadvertent Actuation
a.
Inspection Scope
The inspectors reviewed the plants response to an inadvertent initiation of a deluge
system in the reactor feed pump area of the turbine building on October 22, 2011.
The inspectors walked down affected equipment that was wetted in the area and verified
the plants actions to address potential impacts to indicated reactor power levels were
appropriate. Documents reviewed in this inspection are listed in the Attachment to this
report.
This event follow-up review constituted one sample as defined in IP 71153-05.
b.
Findings
No findings were identified.
28
Enclosure
.2
(Closed) Licensee Event Report (LER) 05000331/2011-002-0: Loss of Ultimate Heat
Sink and Completion of Technical Specification Required Shutdown
a.
Inspection Scope
On August 11, 2011, with the plant operating at full power, the RWS system was
declared inoperable after both intake structure traveling screens became non-functional.
With both RWS subsystems inoperable, the licensee entered TS 3.7.2, Condition B,
which required the reactor to be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Following reactor shutdown, the licensee identified a significant amount of
sand/sediment intrusion in the intake (this event is further discussed in Inspection Report 05000331/2011004). This LER documented four events or conditions subject to an
LER per 10 CFR 50.73. The inspectors verified that the events and conditions reported
in the LER were appropriate, that the safety significance was properly evaluated, and
that corrective actions planned or in place were appropriate to address the causes of the
event. The inspectors did not identify any new or additional issues of concern during
their review of the LER. Documents reviewed are listed in the Attachment to this report.
This LER is closed.
This inspection constituted one event report review sample as defined in IP 71153-05.
b.
Findings
No findings were identified.
4OA5 Other Activities
.1
Operation of an Independent Spent Fuel Storage Installation at Operating Plants
(60855.1)
a.
Inspection Scope
The inspectors observed and evaluated select licensee loading, processing, and transfer
operations of the fifth and sixth canisters during the licensees 2011 dry fuel storage
campaign to verify compliance with the applicable certificate of compliance conditions,
the associated TS, and ISFSI procedures. Specifically, the inspectors observed: heavy
loads practices associated with handling of the Transfer Cask; non-destructive
evaluations of welds on the DSC lid; transfer of the DSC to the ISFSI pad; insertion of
the DSC into a Horizontal Storage Module (HSM); and surveys being performed at the
ISFSI pad.
The inspectors performed tours of the ISFSI pad to assess the material condition of the
pad and HSMs. The inspectors reviewed the licensees evaluations of flammable
materials near the ISFSI and the radiation monitoring program. Additionally, the
inspectors performed independent radiation surveys around the ISFSI pad and HSMs
and verified that the contamination and radiation levels from the Transfer Cask were well
below the regulatory limits.
The inspectors reviewed select documents, in part, after the licensee completed certain
loading activities and a review of the fuel selection documentation was performed to
verify the fuel placed in the DSC met the TS requirements. The inspectors observed the
licensee perform crane operations and reviewed the applicable procedures for
29
Enclosure
compliance with the control of heavy loads program. In addition, the inspectors
reviewed condition reports and the associated corrective actions to verify the licensee
took adequate corrective actions in a timely manner to correct the issues.
The inspectors also reviewed 72.48 screenings and changes to the licensees
10 CFR 72.212 evaluations since the last ISFSI inspection.
b.
Findings
No findings were identified.
4OA6 Management Meetings
.1
Exit Meeting Summary
On January 12, 2012, the inspectors presented the inspection results to Mr. P. Wells,
and other members of the licensee staff. The licensee acknowledged the issues
presented. The inspectors confirmed that none of the potential report input discussed
was considered proprietary.
.2
Interim Exit Meetings
Interim exits were conducted for:
The results of the Emergency Preparedness program inspection were discussed
with Mr. M. Davis via telephone on October 13, 2011.
The ISFSI operational inspection concluded with an interim exit meeting on
November 17, 2011. The inspector presented the inspection results to
Mr. C. Conklin and other members of the licensee management and staff.
Licensee personnel acknowledged the information presented.
Radiological Hazard Assessment and Exposure Controls and Occupational
ALARA Planning and Controls with Mr. P. Wells on December 2, 2011.
The inspectors confirmed that none of the potential report input discussed was
considered proprietary.
Any proprietary material received during the inspection was returned to the licensee.
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, for being dispositioned as an NCV.
The licensee identified a finding of very low safety significance (Green) and an
associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and
Drawings, for the failure to adequately implement post-maintenance testing of the A
intake structure ventilation subsystem temperature controller. Specifically, on
November 7, 2011, maintenance technicians incorrectly performed WO 40039255 for a
replacement temperature controller. The testing was intended to demonstrate
functionality of the A intake structure ventilation subsystem; however, human
performance errors resulted in the test not revealing improper installation issues with the
controller. On November 16, 2011, the licensee identified the performance deficiency
30
Enclosure
and declared the A intake structure ventilation subsystem non-functional and the A
RWS subsystem inoperable, entered the issue into the CAP as CR 1707561, and
restored the A intake structure ventilation subsystem to a functional status.
Because the A RWS subsystem remained available throughout the period of time the
temperature controller was incorrectly installed, reasonable assurance existed to support
the conclusion that the RWS safety function was not impacted.
This failure to meet the requirements of WO 40039255 was a performance deficiency.
The performance deficiency was more than minor because it was associated with the
Mitigating Systems Cornerstone attribute of configuration control and human
performance, and its objective of ensuring the availability, reliability, and capability of
systems that respond to initiating events to prevent undesirable consequences.
Specifically, had the licensee not identified the condition, environmental or operating
conditions could have occurred which could have challenged availability of the A RWS
system or impacted operability. The inspectors determined the finding could be
evaluated using the SDP in accordance with IMC 0609.04, Phase 1 - Initial Screening
and Characterization of Findings, Table 4a. Because the inspectors answered No to
all five questions under Mitigating Systems Cornerstone column, the inspectors
screened the finding as very low safety significance (Green).
ATTACHMENT: SUPPLEMENTAL INFORMATION
1
Attachment
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
P. Wells, Site Vice President
D. Curtland, Plant General Manager
K. Kleinheinz, Site Engineering Director
S. Catron, Licensing Manager
G. Young, Nuclear Oversight Manager
G. Pry, Operations Director
R. Wheaton, Maintenance Site Director
R. Porter, Chemistry & Radiation Protection Manager
B. Kindred, Security Manager
B. Simmons, Training Manager
M. Davis, Emergency Preparedness Manager
B. Murrell, Licensing Engineer Analyst
D. Barta, Licensing Engineer/Analyst
C. Conklin, Project Manager
C. Harberts, Refuel Floor Project Manager
K. Peveler, Nuclear Oversight Supervisor
Nuclear Regulatory Commission
K. Feintuch, Project Manager, NRR
M. Ring, Chief, Reactor Projects Branch 1
2
Attachment
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
Opened 05000331/2011005-01
SLIV
Failure to Make Required Eight Hour Event Report per
10 CFR 50.72(b)(3)(v)(B) (Section 1R15)05000331/2011005-02
Cancellation of River Survey Work Order Causes
Inoperability of River Water Supply System (Section 4OA2.4)05000331/2011005-03
Procedural Non-Compliance Results in Reactor Building
Crane Colliding with ISFSI Inspection Stand
(Section 4OA2.5)
Closed 05000331/2011005-01
SLIV
Failure to Make Required Eight Hour Event Report per
10 CFR 50.72(b)(3)(v)(B) (Section 1R15)05000331/2011005-02
Cancellation of River Survey Work Order Causes
Inoperability of River Water Supply System (Section 4OA2.4)05000331/2011005-03
Procedural Non-Compliance Results in Reactor Building
Crane Colliding with ISFSI Inspection Stand
(Section 4OA2.5)
05000331/2011-002-0
LER
Loss of Ultimate Heat Sink and Completion of Technical
Specification Required Shutdown
Discussed
None.
3
Attachment
LIST OF DOCUMENTS REVIEWED
The following is a partial list of documents reviewed during the inspection. Inclusion on this list
does not imply that the NRC inspector reviewed the documents in their entirety, but rather that
selected sections or portions of the documents were evaluated as part of the overall inspection
effort. Inclusion of a document on this list does not imply NRC acceptance of the document or
any part of it, unless this is stated in the body of the inspection report.
1R01
OP-AA-102-1002 (DAEC); Seasonal Readiness; Revision 5
OP-AA-102-1002; Seasonal Readiness; Revision 0
Abnormal Operating Procedure 903; Severe Weather; Revision 33
OP-AA-102-1002 (DAEC); Seasonal Readiness; Revision 5
CR 01688945; Track Seasonal Readiness Items to Completion Prior to 10/31
DAEC Certification Letter for 2011 Cold Weather Readiness Period
NG-270K; Plant Winterization Checklist; Revision 5
1R04
OI 152A2; HPCI System Valve Lineup and Checklist; Revision 16
OI 152A4; HPCI System Control Panel Lineup; Revision 5
OI 152A1; HPCI System Electrical Lineup; Revision 3
OI 152; High Pressure Coolant Injection System; Revision 99
CR 01706756; While Performing Routine Auxiliary Operator Rounds, a Loud Squeal Was
Discovered on 1VAC014B HPCI Room Cooling Unit
OI 324A10; SBDG Standby/ Readiness Condition Checklist; Revision 14
OI 454A4; B ESW System Valve Lineup and Checklist; Revision13
1R05
ACP 1203.53; Fire Protection; Revision 14
ACP 1412.4; Impairments to Fire Protection Systems; Revision 64
DAEC Fire Plan - Volume 1, Program; Revision 61
AFP 03; Reactor Building HPCI, RCIC & Radwaste Tank Rooms; Revision 26
AFP 07; Reactor Building Laydown Area, Corridor and Waste Tank Area and Spent Resin Tank
Room El. 786; Revision 30
AFP 74; Switchyard; Revision 5
AFP 79; Spent Fuel Storage Facility; Revision 2
AFP -04; Reactor Building North CRD Module Area, CRD repair and CRD Cable Rooms;
Revision 28
AFP-05; Reactor Building South CRD Module Area and Offgas Recombiner Rooms and
Railroad Airlock; Revision 26
AFP-06; Reactor Building RHR Valve Room EL. 757-6; Revision 24
AFP-10; Main Exhaust Fan Room, Heating Hot Water Pump Room and the Plant Air supply Fan
Room; Revision 24
AFP-11; Reactor Building Laydown Area EL. 833-6; Revision 25
AFP-12; Reactor Building Decay Tank and Condensate Phase Separator Rooms; Revision 24
Fire Protection Impairment Permit FPR-09-7320; The DAEC NFPA-805 Transition Project has
Identified Multiple Spurious Operations Vulnerabilities for the RB1 Area
Fire Protection Impairment Permit CMP-11-5115; Dry Fuel Storage Campaign
4
Attachment
AFP 20; Aux Boiler Room, Emergency Diesel Generator Rooms and Generator Day Tank
Rooms El. 7576; Revision 29
1R06
Abnormal Operating Procedure-902; Flood; Revision 39
1R11
Simulator Exercise Guide 2011E-02S; Revision 0
1R12
DAEC Maintenance Rule Program; Cycle 22 Cyclic Report; March 4, 2009 - December 9, 2010
CR 01687326; LIC-1319 Outside Normal Range During STP 3.0.0-02
WO 40112070; LIC-1319 Outside Normal Range During STP 3.0.0-02
CR 01698275; CV-1321, 1E003B Feedwater Heater Dump Valve, Could Not Maintain
Feedwater Heater Level
CR 01698276; Trend - Step Changes in Feedwater Heater 4A and 6B Pressures
CR 01698288; Received 1C06 (D-8) 1E4B Hi-Hi Level Without Receiving the Hi Level
System Level Performance Criteria Basis Document; Feedwater and Condensate; Revision 0
1R13
Work Planning Guideline -1; Work Process Guideline; Revision 47
Work Planning Guideline-2; Online Risk Management Guideline; Revision 59
OP-AA-104-1007; Online Aggregate Risk; Revision 2
WM-AA-1000; Work Activity Risk Management; Revision 11
WM-AA-1000 (DAEC); Work Activity Risk Management (DAEC); Revision 0
OP-AA-102-1003; Guarded Equipment; Revision 3
OP-AA-102-1003 (DAEC); Guarded Equipment (DAEC Specific Information); Revision 19
Work Week 1140 Work Activity Risk Management Summary and Weekly Probabilistic Risk
Analysis
1R15
EN-AA-203-1001; Operability Determinations/ Functionality Assessments; Revision 5
OP-AA-100-1000; Conduct of Operations; Revision 5
TAR 01680290-01; A EDG Exhaust Leak at Opposite Control Side from Cylinders 10, 11, and
12 Exhaust Header Plug
Condition Evaluation (CE) 01680290-02; A EDG Exhaust Leak at Opposite Control Side
Exhaust Header - Cylinders 10, 11, and 12
CR 01700990; Standby Gas Treatment Sump Piping is Corroded
Prompt Operability Determination (POD) 01700990-01; Standby Gas Treatment Sump Piping is
Corroded
OI 710; Intake Structure HVAC System; Revision 15
CR 01718078; OI 710 Provides No Guidance for Intake Unit Heaters Per ARP
Operations Shift Logs; November 7, 2011 through November 17, 2011
CR 01716448; NRC Questions TAR 1707561 Conclusion
5
Attachment
1R19
ACP 1408.1; Work Order Task(s); Revision 169
WO 40075137; Functional Check and Calibration of Transducer
WO 40076624; Oil Circuit Breaker Major Service and Inspection
WO 40079157; Transformer Condition (Diagnostic Electrical Testing)
OI 831.4; Plant Process Computer System (PPC); Revision 74
CR 01699097; LEFM Indicating Bad Inputs
CR 01699182; Degraded LEFM Affecting Indicated Reactor Power
1R22
ACP 107; Surveillance Tests; Revision 13
STP 3.6.1.6-01; Pressure Suppression Chamber to Reactor Building Vacuum Breaker
Operability Test; Revisions 7 and 8
STP 3.8.1-04B; B Standby Diesel Generator Operability Test (Slow Start from Normal Starting
Air); Revision 17
CR 01693756; 1C08B (A-2) B diesel to 1A4 breaker 1A411 Trip Alarm Cycling
CR 01692424; 3.6.1.6-01 - Pressure Suppression Chamber to Reactor Building
STP 3.7.4-01B; B Standby Filter Unit - Logic System Functional Test and Simulated Automatic
Actuation; Revision 2
STP 3.6.1.6-01; Pressure Suppression Chamber to Reactor Building Vacuum Breaker
Operability Test; Revisions 7 and 8
STP 3.8.1-04B; B Standby Diesel Generator Operability Test (Slow Start from Normal Starting
Air); Revision 17
CR 01693756; 1C08B (A-2) B diesel to 1A4 breaker 1A411 Trip Alarm Cycling
CR 01692424; 3.6.1.6-01 - Pressure Suppression Chamber to Reactor Building
1EP4
10 CFR 50.54(q) Evaluation Package; Emergency Plan Implementing Procedure (EPIP) Form
EAL-01; Emergency Action Level Matrix; Revision 8
10 CFR 50.54(q) Evaluation Package; EBD F; Fission Product Barrier Degradation;
Revision 10
10 CFR 50.54(q) Evaluation Package; EBD H; Hazards & Other Conditions Affecting Plant
Safety; Revision 7
10 CFR 50.54(q) Evaluation Package; EPIP 1.2; Notifications; Revision 40
10 CFR 50.54(q) Evaluation Package; EPIP 2.1; Activation and Operation of the Operational
Support Center (OSC); Revision 17
10 CFR 50.54(q) Evaluation Package; EPIP 2.4; Activation and Operation of the ORAA;
Revision 15
10 CFR 50.54(q) Evaluation Package; EPIP 2.8; Security Threat; Revision 8
10 CFR 50.54(q) Evaluation Package; EPIP 3.1; In-Plant Radiological Monitoring; Revision 22
10 CFR 50.54(q) Evaluation Package; EPIP 3.1; In-Plant Radiological Monitoring; Revision 23
10 CFR 50.54(q) Evaluation Package; EPIP 3.2; Field Radiological Monitoring; Revision 19
10 CFR 50.54(q) Evaluation Package; EPIP Form EOF-08; Rad & EOF Manager Checklist;
In-Plant Radiological Monitoring; Revision 13
10 CFR 50.54(q) Evaluation Package; EPIP Form EOF-09; EOF Ops Liaison Checklist;
Revision 9
10 CFR 50.54(q) Evaluation Package; EPIP Form MIDAS-01; MIDAS Operability Test;
Revision 3
6
Attachment
10 CFR 50.54(q) Evaluation Package; EPIP Form NOTE-05; EAL Notification Form;
Revision 13
10 CFR 50.54(q) Evaluation Package; EPIP Form NOTE-07; Basic Notification Flowpath;
Revision 11
10 CFR 50.54(q) Evaluation Package; EPIP Form NOTE-02; ERO Notification - Alphanumeric
Paging System Callout; Revision 8
10 CFR 50.54(q) Evaluation Package; EPIP Form OSC-011; Emergency Assignment Staffing
Board Duties; Revision 4
10 CFR 50.54(q) Evaluation Package; EPIP Form OSC-12; External Exposure Limits;
Revision 1
10 CFR 50.54(q) Evaluation Package; EPIP Form TSC-04; Technical & Engineering Supervisor
Checklist; Revision 7
10 CFR 50.54(q) Evaluation Package; EPIP Form TSC-07; TSC Administrative Supervisor
Checklist; Revision 5
10 CFR 50.54(q) Evaluation Package; EPIP Form TSC-01; TSC Emergency Coordinator
Checklist; Revision 13
10 CFR 50.54(q) Evaluation Package; EPIP Form EOF-07; Emergency Response & Recovery
Director; Revision 13
2RS1
ACP 1411.13; Control of Locked High Radiation Areas and Above; Revision 30
ACP 1411.23; Equipment and Material Controls in Radiological Controlled Areas; Revision 28
ACP 1411.27; Rules for Conduct of Work in Radiologically Controlled Areas; Revision 34
CR 01651746; High Radiation Area Swing-gate Malfunction
CR 01654201; Issues Identified during Low Level Walkdown
CR 01666708; Waste Drum Storage Area Improvements
CR 01666722; Quality Assurance Finding-Oversight of Radioactive Waste Activities
CR 01677279; Sealed Source Leak Test Did Not Meet Schedule Adherence
CR 01685973; Energy Solutions Cask Drive Enters Radiologically Restricted Area Without an
Electronic Dosimeter
CR 01688541; Radiologically Restricted Area High Efficiency Particulate Air Filter Vacuum
Storage Room Inventory Control
HPP 3104.01; Control of Access to High Radiation Areas and Above; Revision 54
HPP3104.09; Drywell Initial Entry and Start-up/Shut-down Entries; Revision 24
HPP 3105.09; Personnel Dosimetry for External Exposure; Revision 26
HPP 3107.04; Radioactive Source Inventory, Control, Leak Check; Revision 16
HPP 3107.05; Release of Items from the Radiologically Restricted Area; Revision 17
RP-AA-100-1001; Radiation Protection Conduct of Operations Guideline; Revision 1
RFO 22; Department Report-Radiation Protection and ALARA; Revision 0
RWP 10-3014; All Cavity Work with the Vessel Filled to the Flange; Revision 00
RWP 10-3016; Pool Work RDO-22; Revision 1
RWP 10-3024; Steam Dryer Tie Bar Replacement; Revision 1
RWP 10-3025; Boron Tube Recovery from Dryer Separator Pit; Revision 00
RWP 10-4213; In-Service Project N2 Penetration Work; Revision 2
RWP 10-4252; Source Range Monitor and Intermediate Range Monitor Removals; Revision 00
RWP 11-22; Management Planning, and Routine Engineering Inquiries; Revision 00
RWP 11-182; Cask Pit Clean-up; Revision 02
RWP 11-206; High Radiation Area/Locked High Radiation Area; Revision 00
RWP 11-249; Dry Fuel Storage Project
WO 0138219; Sealed Source Leakage Test
7
Attachment
WO 40060312; Sealed Source Leakage Test
11 DFS; Dry Fuel Storage Campaign Number 2 ALARA Plan; Revision 0
11-R-004; Reactor Water Clean-up Resin Shipment ALARA Plan; Revision 0
11-1206; Survey Spent Fuel Pool Cask Pit; dated September 12, 2011
11-1065; Survey Drywell 805; dated August 17, 2011
2RS2
ALARA Review 10-N2; Post Job ALARA Review N2A Nozzle Weld Overlay; dated
March 9, 2011
DAEC 5 Year ALARA Plan; Revision 1
10-4163; Survey Drywell 757, BRAC Point Survey; dated November, 19, 2011
10-4164; Survey Drywell 742, BRAC Point Survey; dated November, 19, 2011
4OA1
MSPI Design Basis Document, Revisions 12, 13
NRC PI Data Calculation Review and Approval Packages for MSPI Cooling Water 4th Quarter
2010 through 3rd Quarter 2011
NRC PI Data Calculation Review and Approval Packages for MSPI SSFF 4th Quarter 2010
through 3rd Quarter 2011
NRC PI Data Calculation Review and Approval Packages for MSPI Residual Heat Removal 4th
Quarter 2010 through 3rd Quarter 2011
MSPI Unavailability Index Cooling Water Derivation Report; October 2010 through
September 2011
MSPI Unavailability Index Residual Heat Removal Derivation Report; October 2010 through
September 2011
MSPI Unreliability Index Cooling Water Derivation Report; October 2010 through
September 2011
MSPI Unreliability Index Residual Heat Removal Derivation Report; October 2010 through
September 2011
4OA2
ACP 1410.15; Plant Status Control Program; Revision 6
PI-AA-103-1000; Human Performance Program Error Reduction Tools; Revision 1
ACP 1410.2; LCO Tracking and Safety Function Determination Program; Revision 28
CR 01701934; Cask Service Platform Out of Position
4OA3
CR 01699090; Deluge Initiation Feed Pump Area, Fire Brigade Activated
CR 01699098; Deluge #3 Failed
CR 01699097; LEFM Indicating Bad Inputs
4OA5
ACP 103.4; 10 CFR 72.48 Screening Process; Revision 13
ACP 103.5; 10 CFR 72.48 Evaluation Process; Revision 10
DFS 104; Ancillary Equipment Receipt Inspections and Pre-Op Testing; Revision 5
EC-156669; Reactor Building Crane Trolley Restraints; Revision 3
8
Attachment
DBD-F16-001; Duane Arnold Energy Center Design Basis Document for the Dry Spent Fuel
Storage Program; Revision 11
DAEC-1FJF-11-106; Irradiated Fuel Assembly Selection for Duane Arnold Energy Center 2011
ISFSI Campaign; Revision 3
RFP 403; Performance of Fuel Handling Activities; Revision 45
DFS 203; Dry Shielded Canister Sealing Operations; Revision 27
DFS 301; Loaded Dry Shielded Canister / Transfer Cask from Refueling Floor to ISFSI
Operations; Revision 13
DFS 302; Dry Shielded Canister from Transfer Cask to Horizontal Storage Module Transfer
Operations; Revision 13
VNDR-11-017l; Spent Fuel Cask Welding: 61BT NUHOMS Canisters; Revision 0
7248SCRN-10079; DFS-203 - Dry Shielded Canister Sealing Operations, PCR 01700974;
November 11, 2011
7248SCRN-9830; DFS-203 - Dry Shielded Canister Sealing Operations, PCR 01624499
CR 00343343; HSM Dimensional Tolerances Reduced due to Base Mat Settling
CR 00566670; Additional Actions to Restore Full Qualification of ISFSI
CR 01687477; Cut 480 Cable Supplying the HPU
CR 01700575; Dust on DSC Outer Lid may Interfere with PT of Weld
CR 01700996; NRC ISFSI Inspection Regarding Reactor Building Crane
CR 01701934; Cask Service Platform Out of Position
CR 01703042; Areas for Improvement Identified by NRC ISFSI Inspector
CR 01704968; Inadequate 72.48 Review for Deletion of Helium Leak Test
4OA7
CR 01707561; Work Order Error on Input to TC7715B
WO 40039255; TC7715A Tubing Correction and Operability Testing
9
Attachment
LIST OF ACRONYMS USED
Administrative Control Procedure
Agencywide Document Access Management System
AFP
Area Fire Plan
As-Low-As-Is-Reasonably-Achievable
Annunciator Response Procedure
Corrective Action Program
CFR
Code of Federal Regulations
CR
Condition Report
Duane Arnold Energy Center
Division of Reactor Projects
Dry Storage Canister
Emergency Plan Implementing Procedure
Emergency Service Water
High Pressure Coolant Injection
HSM
Horizontal Storage Module
Heating, Ventilation and Air Conditioning
IMC
Inspection Manual Chapter
IP
Inspection Procedure
IR
Inspection Report
Independent Spent Fuel Storage Installation
Leading Edge Flow Meter
LER
Licensee Event Report
Low Pressure Coolant Injection
Mitigating Systems Performance Index
Non-Cited Violation
NEI
Nuclear Energy Institute
NRC
U.S. Nuclear Regulatory Commission
Operating Instruction
Operational Support Center
Publicly Available Records System
Performance Indicator
Reactor Building
Root Cause Evaluation
Reactor Core Isolation Cooling
RWS
River Water Supply
Significance Determination Process
Severity Level
Surveillance Test Procedure
Technical Assessment for Reportability
Traversing Incore Probe
TS
Technical Specification
Updated Final Safety Analysis Report
Work Order
P. Wells
-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 (PARS) component
of NRC's Agencywide Document Access and Management System (ADAMS). ADAMS is
accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html
(the Public Electronic Reading Room).
Sincerely,
/RA/
Mark Ring, Branch Chief
Branch 1
Division of Reactor Projects
Docket No. 50-331
License No. DPR-49
Enclosure:
Inspection Report 05000331/2011005
w/Attachment: Supplemental Information
cc w/encl:
Distribution via ListServ
DISTRIBUTION:
See next page
DOCUMENT NAME: G:\\DRPIII\\1-Secy\\1-Work In Progress\\DUA 2011 005.docx
Publicly Available
Non-Publicly Available
Sensitive
Non-Sensitive
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
RIII
E RIII
RIII
RIII
NAME
MRing:cs
DATE
01/31/12
OFFICIAL RECORD COPY
Letter to P. Wells from M. Ring dated January 31, 2012
SUBJECT:
DUANE ARNOLD ENERGY CENTER - NRC INTEGRATED INSPECTION
REPORT 05000331/2011005
DISTRIBUTION:
Breeda Reilly
RidsNrrDorlLpl3-1 Resource
RidsNrrPMDuaneArnold Resource
RidsNrrDirsIrib Resource
Cynthia Pederson
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DRPIII
DRSIII
Patricia Buckley
ROPreports.Resource@nrc.gov