ML103020254
| ML103020254 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 10/29/2010 |
| From: | Jamnes Cameron NRC/RGN-III/DRP/B6 |
| To: | Bezilla M FirstEnergy Nuclear Operating Co |
| References | |
| IR-10-004 | |
| Download: ML103020254 (42) | |
See also: IR 05000440/2010004
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION III
2443 WARRENVILLE ROAD, SUITE 210
LISLE, IL 60532-4352
October 29, 2010
Mr. Mark Bezilla
Site Vice President
FirstEnergy Nuclear Operating Company
Perry Nuclear Power Plant
P. O. Box 97, 10 Center Road, A-PY-A290
Perry, OH 44081-0097
SUBJECT: PERRY NUCLEAR POWER PLANT NRC INTEGRATED
INSPECTION REPORT 05000440/2010004
Dear Mr. Bezilla:
On September 30, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed an
inspection at your Perry Nuclear Power Plant. The enclosed report documents the inspection
findings which were discussed on October 6, 2010, with you and 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.
Based on the results of this inspection, two NRC-identified findings of very low safety
significance (Green) were identified. Both of the findings were determined to involve a violation
of NRC requirements, however, because the findings were of very low safety significance and
because the issues were entered into your corrective action program, the NRC is treating the
findings as non-cited violations (NCVs) consistent with Section 2.3.2 of the NRC Enforcement
Policy.
If you contest the subject or severity of these NCVs, you should provide a response
within 30 days of the date of this inspection report, with the basis for your denial, to the
U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington,
DC 20555-0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory
Commission - Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the
Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; and the Resident Inspector Office at the Perry Nuclear Power Plant.
In addition, if you disagree with the cross-cutting aspect of any finding 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 Perry Nuclear Power Plant.
M. Bezilla -2-
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its
enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records (PARS) component of NRC's document system
(ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html
(the Public Electronic Reading Room).
Sincerely,
/RA/
Jamnes L. Cameron, Chief
Branch 6
Division of Reactor Projects
Docket No. 50-440
License No. NPF-58
Enclosure: Inspection Report 05000440/2010004
w/Attachment: Supplemental Information
cc w/encl: Distribution via ListServ
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No: 50-440
License No: NPF-58
Report No: 050000440/2010004
Licensee: FirstEnergy Nuclear Operating Company (FENOC)
Facility: Perry Nuclear Power Plant, Unit 1
Location: Perry, Ohio
Dates: July 1, 2010, through September 30, 2010
Inspectors: M. Marshfield, Senior Resident Inspector
T. Hartman, Resident Inspector
R. Edwards, Reactor Inspector
L. Jones, Reactor Engineer
M. Phalen, Senior Health Physicist, DRS
W. Slawinski, Senior Health Physicist, DRS
P. Smagacz, Reactor Engineer
Observers: V. Myers, Nuclear Safety Professional Development
Program
R. Leidy, Ohio Department of Health
Approved by: Jamnes L. Cameron, Chief
Branch 6
Division of Reactor Projects
Enclosure
TABLE OF CONTENTS
SUMMARY OF FINDINGS ........................................................................................................... 1
REPORT DETAILS ....................................................................................................................... 3
Summary of Plant Status ........................................................................................................... 3
1. REACTOR SAFETY ....................................................................................................... 3
1R01 Adverse Weather Protection (71111.01) ............................................................. 3
1R04 Equipment Alignment (71111.04Q) ..................................................................... 3
1R05 Fire Protection (71111.05Q)................................................................................ 5
1R06 Flood Protection Measures (71111.06) ............................................................... 5
1R11 Licensed Operator Requalification Program (71111.11) ..................................... 6
1R12 Maintenance Effectiveness (71111.12Q) ............................................................ 6
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13) ......... 7
1R15 Operability Evaluations (71111.15) ..................................................................... 8
1R18 Temporary Plant Modifications (71111.18) ......................................................... 9
1R19 Post-Maintenance Testing (71111.19) ................................................................ 9
1R22 Surveillance Testing (71111.22) ....................................................................... 10
2. RADIATION SAFETY ................................................................................................... 13
2RS1 Radiological Hazard Assessment and Exposure Controls (71124.01) ............. 13
2RS3 In-Plant Airborne Radioactivity Control and Mitigation (71124.03) ................... 17
2RS4 Occupational Dose Assessment (71124.04) ..................................................... 20
4. OTHER ACTIVITIES ..................................................................................................... 21
4OA1 Performance Indicator Verification (71151)....................................................... 21
4OA2 Problem Identification and Resolution (71152) ................................................. 23
4OA3 Follow-up of Events and Notices of Enforcement Discretion (71153) ............... 26
4OA5 Other Activities .................................................................................................. 27
4OA6 Meetings............................................................................................................ 28
4OA7 Licensee-Identified Violations ........................................................................... 28
SUPPLEMENTAL INFORMATION ............................................................................................... 1
Key Points of Contact ................................................................................................................ 1
List of Items Opened, Closed and Discussed............................................................................ 1
List of Documents Reviewed ..................................................................................................... 2
List of Acronyms Used .............................................................................................................. 8
Enclosure
SUMMARY OF FINDINGS
IR 05000440/2010004; 07/01/2010 - 09/30/2010; Surveillance Testing; Problem Identification
and Resolution.
The inspection was conducted by resident and regional inspectors. The inspection report
(IR) covers a 3-month period of resident inspection. Two green findings which were NCVs were
identified. The significance of most findings is indicated by their color (Green, White, Yellow, or
Red) using Inspection Manual Chapter (IMC) 0609 Significance Determination Process (SDP).
Cross-cutting aspects were determined using IMC 0310, "Components Within The Cross-
Cutting Areas." Findings for which the SDP does not apply may be "Green," or be assigned a
severity level after NRC management review. The NRC's 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. Inspector-Identified and Self-Revealed Findings
Cornerstone: Mitigating Systems
- Green. The inspectors identified a finding of very low safety significance and associated
NCV of 10 CFR 50, Appendix B, Criterion XI, Test Control, for the unacceptable
preconditioning of the 'A' residual heat removal (RHR) pump minimum flow valve prior to
quarterly in-service testing. Specifically, the licensee performed a surveillance that
cycled the valve prior to performing stroke time testing, and had not previously
performed an evaluation assessing the sequence for preconditioning. The licensee
entered the issue into their corrective action program.
The inspectors determined that unacceptably preconditioning the RHR minimum flow
valve was a performance deficiency that affected the Mitigating Systems Cornerstone
because it can mask the true as-found condition of a component designed to mitigate
accidents. The performance deficiency was determined to be more than minor because,
if left uncorrected, it could lead to a more significant safety concern. The finding was of
very low safety significance because it was not a design/qualification deficiency, did not
represent a loss of system safety function, did not result in a loss of function of a single
train for greater than its Technical Specification (TS)-allowable outage time, did not
result in a loss of function of nonsafety-related risk-significant equipment and was not
risk significant due to external events. This finding has a cross-cutting aspect in the
work control planning component of the Human Performance area (per
IMC 0310 H.3(a)), because the licensee did not appropriately plan work activities for
plant structures, systems, and components. Specifically, the licensee did not schedule
the surveillance tests in the proper sequence to prevent unacceptable preconditioning of
the valve. (Section 1R22)
- Green. The inspectors identified a finding of very low safety significance and associated
NCV for a failure to comply with TS 3.0.2 by not entering TS Limiting Condition for
Operation (LCO) 3.3.5.1 Condition A and TS LCO 3.3.6.1 Condition A when required.
The inspectors determined that the licensee incorrectly utilized a TS Surveillance
Requirement Note that allows a delay in entering the Conditions and Required Actions
for the given TS LCO. As a result, the licensee failed to correctly enter the Conditions
and Required Actions when reactor level instruments were declared inoperable to
1 Enclosure
perform testing in support of planned maintenance. The licensee entered the issue
associated with the failure to comply with TS into their corrective action program.
This performance deficiency was determined to be more than minor because it impacted
the Equipment Performance attribute of the Mitigating Systems Cornerstone, and
adversely affected the cornerstone objective to ensure the availability, reliability, and
capability of systems that respond to initiating events to prevent undesirable
consequences (i.e., core damage); and if left uncorrected it could lead to a more
significant safety concern. This finding is of very low safety significance because it was
not a design/qualification deficiency, did not represent a loss of system safety function,
did not result in a loss of function of a single train for greater than its TS-allowable
outage time, did not result in a loss of function of nonsafety-related risk-significant
equipment and was not risk significant due to external events. This finding has a
cross-cutting aspect in the decision making component of Human Performance
cross-cutting area (per IMC 0310 H.1(a)), because the licensee did not use conservative
assumptions to ensure the proposed action was safe. Specifically, the licensee
incorrectly used the TS Surveillance Requirement Note to satisfy maintenance
requirements. (Section 4OA2)
B. Licensee-Identified Violations
One violation of very low safety significance was identified by the licensee and
has been reviewed by the inspectors. Corrective actions taken or planned by the
licensee have been entered into the licensee's corrective action program. This
violation and its corrective action tracking number are listed in Section 4OA7 of
this report.
2 Enclosure
REPORT DETAILS
Summary of Plant Status
The plant began the inspection period at 100 percent power. With the exception of minor
reductions in power to support routine surveillances and rod pattern adjustments, and several
occasions when the plant reduced power because of plant cooling limitations caused by
summer environmental conditions, the plant remained at full power for the entire period.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and
1R01 Adverse Weather Protection (71111.01)
External Flooding
a. Inspection Scope
The inspectors evaluated the design, material condition, and procedures for coping with
the design basis probable maximum flood. The evaluation included a review to check
for deviations from the descriptions provided in the Updated Safety Analysis Report
(USAR) for features intended to mitigate the potential for flooding from external factors.
As part of this evaluation, the inspectors checked for obstructions that could prevent
draining, checked that the roofs did not contain obvious loose items that could clog
drains in the event of heavy precipitation, and determined that barriers required to
mitigate the flood were in place and operable. Additionally, the inspectors performed a
walkdown of the protected area to identify any modification to the site which would inhibit
site drainage during a probable maximum precipitation event or allow water ingress past
a barrier. The inspectors walked down underground bunkers/manholes subject to
flooding that contained multiple train or multi-function risk-significant cables. The
inspectors also reviewed the Off-Normal Instructions (ONIs) for mitigating the design
basis flood to ensure it could be implemented as written.
This inspection constituted one sample of external flooding as defined in Inspection
Procedure (IP) 71111.01-05.
b. Findings
No findings were identified.
1R04 Equipment Alignment (71111.04Q)
a. Inspection Scope
The inspectors performed partial system walkdowns of the following risk-significant
systems:
- 'A' motor control center, switchgear and miscellaneous electrical equipment
heating ventilation and air conditioning system on September 2, 2010; and
3 Enclosure
- 'B' reactor protection system (RPS) power supply electrical alignment while 'A'
RPS motor generator set was out of service on September 30, 2010.
The inspectors selected these systems based on their risk-significance relative to the
Reactor Safety Cornerstone 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, USAR, 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 corrective action program (CAP) with the appropriate significance
characterization. Documents reviewed are listed in the Attachment.
These inspections constituted three partial system walkdown samples for equipment
alignment as defined in IP 71111.04-05.
b. Findings
No findings were identified.
.2 Semi-Annual Complete System Walkdown
a. Inspection Scope
On September 24, 2010, the inspectors concluded a complete system alignment
inspection of the emergency closed cooling (ECC) system to verify the functional
capability of the system. This system was selected because it was considered both
safety significant and risk significant in the licensees probabilistic risk assessment.
The inspectors walked down the system to review mechanical and electrical equipment
line-ups, electrical power availability, system temperature indications, component
labeling, component lubrication, component and equipment cooling, hangers and
supports, operability of support systems, and to ensure that ancillary equipment or
debris did not interfere with equipment operation. A review of a sample of past and
outstanding WOs was performed to determine whether any deficiencies significantly
affected the system function. In addition, the inspectors reviewed the CAP database to
ensure that system equipment alignment problems were being identified and
appropriately resolved. Documents reviewed are listed in the Attachment.
This inspection constituted one complete system walkdown sample as defined in
IP 71111.04-05.
b. Findings
No findings were identified.
4 Enclosure
1R05 Fire Protection (71111.05Q)
a. Inspection Scope
The inspectors conducted fire protection walkdowns which were focused on availability,
accessibility, and the condition of firefighting equipment in the following risk-significant
plant areas:
- Fire Zone 0IB-4; Intermediate Building 654'-6" Elevation;
- Fire Zone 0IB-3; Intermediate Bldg 620' Elevation;
- Fire Zone 0CC-2; Control Complex 599' Elevation;
- Fire Zone 0IB-1; Intermediate Bldg 574' Elevation; and
- Fire Zone 1AB-3B; Auxiliary Building 620'-6" Elevation West.
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 inoperable 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, the inspectors verified that fire hoses and
extinguishers were in their designated locations and available for immediate use; that
fire detectors and sprinklers were unobstructed; that transient material loading was
within the analyzed limits; and fire doors, dampers, and penetration seals appeared to
be in satisfactory condition. The inspectors also verified that minor issues identified
during the inspection were entered into the licensees CAP. Documents reviewed are
listed in the Attachment to this report.
These activities constituted five quarterly fire protection inspection samples as defined in
IP 71111.05-05.
b. Findings
No findings were identified.
1R06 Flood Protection Measures (71111.06)
a. Inspection Scope
The inspectors reviewed selected risk important plant design features and licensee
procedures intended to protect the plant and its safety-related equipment from internal
flooding events. The inspectors reviewed flood analyses and design documents,
including the USAR, engineering calculations, and ONI's to identify licensee
commitments. The specific documents reviewed are listed in the Attachment to this
report. In addition, the inspectors reviewed licensee drawings to identify areas and
equipment that may be affected by internal flooding caused by the failure or
misalignment of nearby sources of water, such as the fire suppression or the circulating
5 Enclosure
water systems. The inspectors also reviewed the licensees corrective action documents
with respect to past flood-related items identified in the CAP to verify the adequacy of
the corrective actions.
The inspectors performed a walkdown of the low pressure core spray areas to assess
the adequacy of watertight doors and verify drains and sumps were clear of debris and
were operable, and that the licensee complied with its commitments.
This inspection constituted one internal flooding sample as defined in IP 71111.06-05.
b. Findings
No findings were identified.
1R11 Licensed Operator Requalification Program (71111.11)
a. Inspection Scope
On August 30, 2010, 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;
- 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
- the 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 operator license requalification program sample
as defined in IP 71111.11.
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness (71111.12Q)
a. Inspection Scope
The inspectors evaluated degraded performance issues involving the following
risk-significant systems:
6 Enclosure
- 'B' compressible gas mixing compressor;
- 'A' control room ventilation system;
- Division 3 emergency diesel generator (EDG) and high pressure core spray
(HPCS) system; and
- Upper and lower containment airlocks.
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.
This inspection constituted four 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:
- conservative grid operations on July 15, 2010;
- work on control rod drive pump 'B' concurrent with testing of the Division 3 EDG
on August 17, 2010;
- EDG fuel oil samples on August 26, 2010;
- reactor feed booster pump discharge check valve repair during the week of
September 13, 2010; and
- HPCS diesel generator repairs during the week of September 15, 2010.
7 Enclosure
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.
These maintenance risk assessments and emergent work control activities constituted
five samples as defined in IP 71111.13-05.
b. Findings
No findings were identified.
1R15 Operability Evaluations (71111.15)
a. Inspection Scope
The inspectors reviewed the following issues:
- Technical Support Center ventilation with degraded flow turning vanes;
- EDG ventilation systems with installed relays beyond the 20-year qualification
life;
- Emergency Closed Cooling cross-connect to fuel pool cooling and cleanup valve
failed stroke time testing; and
- 'A' control room ventilation plenum missing insulation.
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 USAR to the licensees evaluations to determine
whether the components or systems were operable. Where compensatory measures
were required to maintain operability, the inspectors determined whether the measures
in place would function as intended and were properly controlled. The inspectors
determined, where appropriate, compliance with bounding limitations associated with the
evaluations. Additionally, the inspectors reviewed a sampling of corrective action
documents to verify that the licensee was identifying and correcting any deficiencies
associated with operability evaluations. Documents reviewed are listed in the
Attachment to this report.
This operability inspection constituted four samples as defined in IP 71111.15-05.
b. Findings
No findings were identified.
8 Enclosure
1R18 Temporary Plant Modifications (71111.18)
a. Inspection Scope
The inspectors reviewed the temporary modification for the Hot Surge Tank Hi/Low
Level Alarm. The inspectors compared the temporary configuration changes and
associated 10 CFR 50.59 screening and evaluation information against the design basis,
the USAR, and the TS, as applicable, to verify that the modification did not affect the
operability or availability of the affected system(s). The inspectors also compared the
licensees information to operating experience information to ensure that lessons learned
from other utilities had been incorporated into the licensees decision to implement the
temporary modification. The inspectors, as applicable, performed field verifications to
ensure that the modifications were installed as directed; the modifications operated as
expected; modification testing adequately demonstrated continued system operability,
availability, and reliability; and that operation of the modifications did not impact the
operability of any interfacing systems. Lastly, the inspectors discussed the temporary
modification with operations, engineering, and training personnel to ensure that the
individuals were aware of how extended operation with the temporary modification in
place could impact overall plant performance. Documents reviewed in the course of this
inspection are listed in the Attachment to this report.
This inspection constituted one sample of a temporary modification as defined in
IP 71111.18-05.
b. Findings
No findings were identified.
1R19 Post-Maintenance Testing (71111.19)
a. Inspection Scope
The inspectors reviewed the following post-maintenance (PM) activities to verify that
procedures and test activities were adequate to ensure system operability and functional
capability:
- safety relief valve control switch replacement during the week of August 2, 2010;
- high drywell pressure master trip unit replacement during the week of
August 9, 2010;
- AEGTS fan replacement during the week of August 25, 2010;
- emergency service water (ESW) ventilation system outlet damper hydramotor
work during the week of September 7, 2010;
- HPCS pump breaker relay replacement during the week of September 17, 2010;
and
- Division 3 EDG outage retest during the week of September 24, 2010.
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;
9 Enclosure
- 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 TS, the USAR, 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 PM tests to determine whether the licensee was identifying problems
and entering them in the CAP and that the problems were being corrected
commensurate with their importance to safety. Documents reviewed are listed in the
Attachment to this report.
This inspection constituted six PM testing samples as defined in IP 71111.19-05.
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:
- Residual Heat Removal (RHR) 'A' pump and valve inservice testing during the
week of July 12, 2010 (IST);
- Emergency Service Water (ESW) 'C' pump and valve operability test during the
week of July 23, 2010 (routine);
- Reactor Core Isolation Cooling (RCIC) pump and valve operability test during the
week of August 2, 2010 (routine); and
- ESW 'B' pump and valve operability testing during the week of August 13, 2010
(routine).
The inspectors observed in-plant activities and reviewed procedures and associated
records to determine the following:
- 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;
10 Enclosure
- as-left setpoints were within required ranges, and the calibration frequency were
in accordance with TS, the USAR, 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 IST activities, testing was performed in accordance with the
applicable version of Section XI, American Society of Mechanical Engineers
(ASME) 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 three routine surveillance testing samples and one inservice
testing sample as defined in IP 71111.22, Sections -02 and -05.
b. Findings
Introduction: The inspectors identified a finding of very low safety significance (Green)
and associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for
unacceptable preconditioning of the 'A' RHR pump minimum flow valve prior to quarterly
IST. Specifically, the surveillance test sequencing caused this valve to be opened and
closed before the documented stroke time testing, and the sequence had not been
evaluated for preconditioning prior to performance of the tests.
Description: On July 8, 2010, at approximately 9:30 a.m., the inspectors observed the
performance of surveillance test SVI-E12-T2001, RHR A Pump and Valve Operability
Test. Included in this test is the quarterly timed valve stroke of 1E12-F0064A, RHR
Pump A Min Flow Valve, as required by the IST program. During review of the previous
shift narrative logs, it was identified that surveillance test SVI-E12-T1194, LPCI (Low
Pressure Core Injection) Pump A Discharge Low Flow (Bypass) Channel Functional for
1E12-N652A, was performed at around 1:30 a.m. This surveillance calibrates
11 Enclosure
instrument 1E12-N652A, LPCI Pump A Discharge Low Flow Instrument. The calibration
of the low flow instrument results in the 'A' train RHR pump minimum flow valve stroking.
This sequence of testing fully cycled the valve several times less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to
obtaining the IST stroke timing data during SVI-E12-T2001.
Inspection Manual Technical Guidance 9900 defines unacceptable preconditioning, in
part, as The alteration; variation; manipulation; or adjustment of the physical condition
of a structure, system, and component (SSC) before or during TS surveillance or ASME
Code testing that will alter one or more of an SSCs operational parameters, which
results in acceptable test results. Such changes could mask the actual as-found
condition of the SSC and possibly result in an inability to verify the operability of the
SSC. In addition, unacceptable preconditioning could make it difficult to determine
whether the SSC would perform its intended function during an event in which the SSC
might be needed. Technical Guidance 9900 further describes that some types of
preconditioning may be considered acceptable, but that this preconditioning should
have been evaluated and documented in advance of the surveillance. Since the
licensee had not performed an evaluation which justified that preconditioning of the valve
was acceptable prior to completing the testing, the licensees surveillance testing
sequence that cycled the valve prior to obtaining stroke time data constituted
unacceptable preconditioning of the valve.
Additionally, the unacceptable preconditioning of the RHR valve was not in accordance
with the licensees procedural guidance regarding IST. Licensee Nuclear Operating
Procedure (NOP)-ER-3204, Inservice Testing Program, states, in part, Maintenance
activities should not be scheduled to influence the results of upcoming tests. Such
actions, known as preconditioning, should be avoided. In addition it also states, in part,
Care should be taken to ensure that procedures, surveillances, or tasks are not
scheduled such that unacceptable preconditioning of a component prior to the inservice
test occurs. Where unacceptable preconditioning would occur, the procedure/task
should specify that an as found test be performed first.
The licensee performed an investigation which revealed that, historically, these two
surveillances had been completed in either sequence without significant differences in
measured stroke time. As a result, the licensee determined that the preconditioning did
not significantly affect the IST stroke timing of the valve. The licensee used this
information to support an operability declaration for the system.
Analysis: The inspectors determined that stroking of the RHR minimum flow valve prior
to as-found stroke timing constituted unacceptable preconditioning and a performance
deficiency. Specifically, performing the IST surveillance test in this sequence may not
accurately indicate potential valve degradation. The inspectors determined that the
performance deficiency affected the Mitigating Systems Cornerstone, because it could
mask the true as-found condition of a component designed to mitigate accidents. The
inspectors evaluated the performance deficiency in accordance with Inspection Manual
Change (IMC) 0612, Appendix B, Issue Screening. This performance deficiency was
compared to, and was not similar to any of, the examples in IMC 0612, Appendix E,
Examples of Minor Issues, but was characterized as more than minor because, if left
uncorrected, it could lead to a more significant safety concern.
The inspectors determined the finding could be evaluated using the SDP in
accordance with IMC 0609, Significance Determination Process, Attachment 0609.04,
12 Enclosure
Phase 1 - Initial Screening and Characterization of Findings, Table 3b for the Mitigating
Systems Cornerstone. The inspectors determined the finding was of very low risk
significance because it was not a design/qualification deficiency, did not represent a loss
of system safety function, did not result in a loss of function of a single train for greater
than its TS allowable outage time, did not result in a loss of function of nonsafety-related
risk-significant equipment, and was not risk significant due to external events.
This finding has a cross-cutting aspect in the work control planning component of the
Human Performance cross-cutting area (per IMC 0310 H.3(a)), because the licensee did
not appropriately plan work activities for plant SSCs and components. Specifically, the
licensee did not schedule the surveillance tests in the proper sequence to prevent
unacceptable preconditioning of the valve.
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion XI, Test Control, states, in
part, that A test program shall be established to assure that all testing required to
demonstrate that structures, systems, and components will perform satisfactorily in
service is identified and performed in accordance with written test procedures which
incorporate the requirements and acceptance limits contained in applicable design
documents. Contrary to this requirement, on July 8, 2010, the licensee stroked
1E12-F0064A, RHR Pump A Min Flow Valve for test procedure SVI-E12-T1194 prior to
performing IST stroke timing, and failed to prevent unacceptable pre-conditioning of the
pump minimum flow valve. Because this finding is of very low safety significance and
because it was entered into the licensees CAP as CR 10-79624, this violation is being
treated as an Non-Cited Violation (NCV) consistent with Section 2.3.2 of the NRC
Enforcement Policy. (NCV 05000440/2010004-01; Unacceptable Preconditioning of
RHR Valve Prior to ASME In-Service Testing.)
2. RADIATION SAFETY
Cornerstones: Public and Occupational Radiation Safety
2RS1 Radiological Hazard Assessment and Exposure Controls (71124.01)
The inspection activities supplement those documented in Inspection Report (IR) 05000440/2010003, and constitute one complete sample as defined in
IP 71124.01-05.
.1 Radiological Hazard Assessment (02.02)
a. Inspection Scope
The inspectors determined if there had been changes to plant operations since the last
inspection that could result 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 these changes and has implemented periodic monitoring, as
appropriate, to detect and quantify the radiological hazard.
The inspectors reviewed the last two radiological surveys from selected plant areas.
The inspectors evaluated whether the thoroughness and frequency of the surveys were
appropriate for the given radiological hazards.
13 Enclosure
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 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 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.
.2 Instructions to Workers (02.03)
a. Inspection Scope
The inspectors selected various containers holding nonexempt licensed radioactive
materials that may cause 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).
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.
.3 Contamination and Radioactive Material Control (02.04)
a. Inspection Scope
The inspectors observed several locations where the licensee monitors potentially
contaminated material leaving the radiologically controlled area and evaluated the
methods used for the control, survey, and release of materials from these areas. The
inspectors also observed the performance of personnel surveying and releasing material
for unrestricted use to determine if the methods used were in accordance with
procedures and whether those procedures were sufficient to control the spread of
contamination and prevent unintended release of materials from the site. The inspectors
determined whether radiation monitoring instrumentation used for these surveys had
appropriate sensitivity for the types of radiation present.
The inspectors reviewed the licensees criteria for the survey and release of potentially
contaminated material to determine if there was guidance on how to respond to an alarm
that indicates 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
14 Enclosure
appropriate counting parameters. The inspectors assessed whether or not the licensee
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 three sealed sources from the licensees inventory records and
assessed whether the sources were accounted for and verified to be intact (i.e., they
were not leaking their radioactive content).
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.
.4 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
(RWPs), and worker briefings. The inspectors reviewed RWPs for work within airborne
radioactivity areas with the potential for individual worker internal exposures. For these
RWPs, 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 (HEPA)
ventilation system operation for selected airborne radioactive material areas
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 (HRAs) and very-high-radiation areas (VHRAs) to verify conformance with the
occupational performance indicator (PI).
b. Findings
No findings were identified.
.5 Risk-Significant High-Radiation Area and Very High-Radiation Area Controls (02.06)
a. Inspection Scope
The inspectors discussed with the radiation protection (RP) manager the controls and
procedures for HRAs and VHRAs. The inspectors discussed methods employed by the
licensee to provide stricter control of VHRA access as specified in 10 CFR 20.1602,
15 Enclosure
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 level of worker protection.
The inspectors discussed the controls in place for special areas that have the potential
to become VHRAs during certain plant operations with first-line health physics (HP)
supervisors (or equivalent positions having backshift HP oversight authority). The
inspectors assessed whether these plant operations required communication
beforehand with the HP group, so as to allow corresponding timely actions to properly
post, control, and monitor the radiation hazards including re-access authorization.
The inspectors evaluated licensee controls for VHRAs and areas with the potential to
become a VHRA to ensure that an individual was not able to gain unauthorized access
to the VHRA.
b. Findings
No findings were identified.
.6 Radiation Worker Performance (02.07)
a. Inspection Scope
The inspectors observed radiation worker performance with respect to stated RP work
requirements. The inspectors assessed whether workers were aware of the radiological
conditions in their workplace and the RWP controls/limits in place, and whether their
performance reflected the level of radiological hazards present.
The inspectors reviewed a maximum of 10 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 RP manager any problems with the corrective actions planned or
taken.
b. Findings
No findings were identified.
.7 Radiation Protection Technician Proficiency (02.08)
a. Inspection Scope
The inspectors observed the performance of the RP technicians with respect to all RP
work requirements. The inspectors evaluated whether technicians were aware of the
radiological conditions in their workplace and the RWP controls/limits, and whether their
performance was consistent with their training and qualifications with respect to the
radiological hazards and work activities.
16 Enclosure
The inspectors reviewed a maximum of 10 radiological problem reports since the last
inspection that found the cause of the event to be RP 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.
2RS3 In-Plant Airborne Radioactivity Control and Mitigation (71124.03)
The inspection activities supplement those documented in IR 05000440/2010003, and
constitute one complete sample as defined in IP 71124.03-05.
.1 Inspection Planning (02.01)
a. Inspection Scope
The inspectors reviewed the plant USAR to identify areas of the plant designed as
potential airborne radiation areas and any associated ventilation systems or airborne
monitoring instrumentation. Instrumentation review included continuous air monitors
(continuous air monitors and particulate-iodine-noble-gas-type instruments) used to
identify changing airborne radiological conditions such that actions to prevent an
overexposure may be taken. The review included an overview of the respiratory
protection program and a description of the types of devices used.
The inspectors reviewed USAR, TS, and emergency planning documents to identify
location and quantity of respiratory protection devices stored for emergency use.
The inspectors reviewed the licensees procedures for maintenance, inspection, and use
of respiratory protection equipment including self-contained breathing apparatus
(SCBA). Additionally, the inspectors reviewed procedures for air quality maintenance
and the reported PIs to identify any related to unintended dose resulting from intakes of
radioactive materials.
b. Findings
No findings were identified.
.2 Engineering Controls (02.02)
a. Inspection Scope
The inspectors reviewed the licensees use of permanent and temporary ventilation to
determine whether the licensee used ventilation systems as part of its engineering
controls (in lieu of respiratory protection devices) to control airborne radioactivity. The
inspectors reviewed procedural guidance for use of installed plant systems, such as
containment purge, spent fuel pool ventilation, and auxiliary building ventilation, and
assessed whether the systems are used, to the extent practicable, during high-risk
activities (e.g., using containment purge during cavity flood up).
17 Enclosure
The inspectors selected installed ventilation systems used to mitigate the potential for
airborne radioactivity, and evaluated whether the ventilation airflow capacity, flow path
(including the alignment of the suction and discharges), and filter/charcoal unit
efficiencies, as appropriate, were consistent with maintaining concentrations of airborne
radioactivity in work areas below the concentrations of an airborne area to the extent
practicable.
The inspectors selected temporary ventilation system setups (HEPA/charcoal negative
pressure units, down draft tables, tents, metal Kelly buildings, and other enclosures)
used to support work in contaminated areas. The inspectors assessed whether the use
of these systems was consistent with licensee procedural guidance and as-low-as-is-
reasonably-achievable (ALARA) concepts.
b. Findings
No findings were identified.
.3 Use of Respiratory Protection Devices (02.03)
a. Inspection Scope
For those situations where it is impractical to employ engineering controls to minimize
airborne radioactivity, the inspectors assessed whether the licensee provided respiratory
protective devices such that occupational doses are ALARA. The inspectors selected
work activities where respiratory protection devices were used to limit the intake of
radioactive materials, and assessed whether the licensee performed an evaluation
concluding that further engineering controls were not practical and that the use of
respirators was ALARA. The inspectors also evaluated whether the licensee had
established means (such as routine bioassay) to determine if the level of protection
(protection factor) provided by the respiratory protection devices during use was at least
as good as that assumed in the licensees work controls and dose assessment.
The inspectors assessed whether respiratory protection devices used to limit the intake
of radioactive materials were certified by the National Institute for Occupational Safety
and Health/Mine Safety and Health Administration (NIOSH/MSHA) or have been
approved by the NRC in accordance with 10 CFR 20.1703(b). The inspectors selected
work activities where respiratory protection devices were used. The inspectors
evaluated whether the devices were used consistent with their NIOSH/MSHA
certification or any conditions of their NRC approval.
The inspectors reviewed records of air testing for supplied-air devices and SCBA bottles
to assess whether the air used in these devices meets or exceeds Grade D quality. The
inspectors reviewed plant breathing air supply systems to determine whether they meet
the minimum pressure and airflow requirements for the devices in use.
The inspectors selected individuals qualified to use respiratory protection devices, and
assessed whether they have been deemed fit to use the devices by a physician.
The inspectors selected several individuals assigned to wear a respiratory protection
device and observed them donning, doffing, and functionally checking the device as
appropriate. Through interviews with these individuals, the inspectors evaluated
whether they knew how to safely use the device and how to properly respond to any
18 Enclosure
device malfunction or unusual occurrence (loss of power, loss of air, etc.). The
inspectors reviewed training curricula for users of the devices.
The inspectors chose various respiratory protection devices staged and ready for use in
the plant or stocked for issuance. The inspectors assessed the physical condition of the
device components (mask or hood, harnesses, air lines, regulators, air bottles, etc.) and
reviewed records of routine inspection for each. The inspectors selected several of the
devices and reviewed records of maintenance on the vital components (e.g., pressure
regulators, inhalation/exhalation valves, hose couplings). The inspectors assessed
whether onsite personnel assigned to repair vital components had received vendor-
provided training.
b. Findings
No findings were identified.
.4 Self-Contained Breathing Apparatus for Emergency Use (02.04)
a. Inspection Scope
Based on USAR, TS, and emergency operating procedure requirements, the inspectors
reviewed the status and surveillance records of SCBAs staged in-plant for use during
emergencies. The inspectors reviewed the licensees capability for refilling and
transporting SCBA air bottles to and from the control room and operations support
center during emergency conditions.
The inspectors selected individuals on control room shift crews, and individuals from
designated departments currently assigned emergency duties (e.g., onsite search and
rescue duties) to assess whether control room operators and other emergency response
and RP personnel (assigned in-plant search and rescue duties or as required by
emergency operating procedures or the emergency plan) were trained and qualified in
the use of SCBAs (including personal bottle change out). The inspectors evaluated
whether personnel assigned to refill bottles were trained and qualified for that task.
The inspectors determined whether appropriate mask sizes and types were available for
use (i.e., in-field mask size and type matched what was used in fit-testing). The
inspectors selected various on-shift operators to determine whether they have no facial
hair that would interfere with the sealing of the mask to the face and whether vision
correction (e.g., glasses inserts or corrected lenses) were available as appropriate.
The inspectors reviewed the past 2 years of maintenance records for several SCBA
units used to support operator activities during accident conditions and designated as
ready for service to assess whether any maintenance or repairs on any SCBA units
vital components were performed by an individual, or individuals, certified by the
manufacturer of the device to perform the work. The vital components typically are the
pressure-demand air regulator and the low-pressure alarm. The inspectors reviewed the
onsite maintenance procedures governing vital component work to determine any
inconsistencies with the SCBA manufacturers recommended practices. For those
SCBAs designated as ready for service, the inspectors determined whether the
required, periodic air cylinder hydrostatic testing was documented and up to date, and
the retest air cylinder markings required by the U.S. Department of Transportation were
in place.
19 Enclosure
b. Findings
No findings were identified.
.5 Problem identification and Resolution (02.05)
a. Inspection Scope
The inspectors reviewed CRs and other corrective action documents to determine
whether problems associated with control and mitigation of in-plant airborne radioactivity
were being identified at the appropriate threshold and were properly addressed for
resolution in the licensees CAP.
b. Findings
No findings were identified
2RS4 Occupational Dose Assessment (71124.04)
This inspection constituted a partial sample as defined in IP 71124.04-05.
.1 Inspection Planning (02.01)
a. Inspection Scope
The inspectors reviewed the results of RP program audits related to internal and external
dosimetry (e.g., licensees quality assurance audits, self-assessments, or other
independent audits) to gain insights into overall licensee performance in the area of dose
assessment and focus the inspection activities consistent with the principle of smart
sampling.
b. Findings
No findings were identified.
.2 Internal Dosimetry (02.03)
Internal Dose Assessment - Airborne Monitoring
a. Inspection Scope
The inspectors reviewed the licensee's program for airborne radioactivity assessment
and dose assessment, as applicable, based on airborne monitoring and calculations of
derived air concentration. The inspectors determined whether flow rates and collection
times for air sampling equipment were adequate to allow lower limits of detection to be
obtained. The inspectors also reviewed the adequacy of procedural guidance to assess
internal dose if respiratory protection was used. The licensee had not performed dose
assessments using airborne/derived air concentration monitoring since the last
inspection.
20 Enclosure
b. Findings
No findings were identified.
.3 Special Dosimetric Situations (02.04)
Dosimeter Placement and Assessment of Effective Dose Equivalent for External
Exposures.
a. Inspection Scope
The inspectors reviewed the licensee's methodology for monitoring external dose in
non-uniform radiation fields or where large dose gradients exist. The inspectors
evaluated the licensee's criteria for determining when alternate monitoring, such as use
of multi-badging, was to be implemented.
The inspectors reviewed dose assessments performed using multi-badging to evaluate
whether the assessment was performed consistent with licensee procedures and
dosimetric standards.
b. Findings
No findings were identified.
4. OTHER ACTIVITIES
4OA1 Performance Indicator Verification (71151)
.1 Mitigating Systems Performance Index - Heat Removal System
a. Inspection Scope
The inspectors sampled licensee submittals for the Mitigating Systems Performance
Index (MSPI) - Heat Removal System performance indicator for the period from the third
quarter 2009 through the second quarter 2010. To determine the accuracy of the PI
data reported during those periods, PI definitions and guidance contained in the Nuclear
Energy Institute (NEI) Document 99-02, Regulatory Assessment Performance Indicator
Guideline, Revision 6, dated October 2009, were used. The inspectors reviewed the
licensees operator narrative logs, issue reports, event reports, MSPI derivation reports,
and NRC Integrated Inspection Reports for the period of the third quarter 2009 through
the second quarter 2010 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
issue report database to determine if any problems had been identified with the PI data
collected or transmitted for this indicator and none were identified. Documents reviewed
are listed in the Attachment to this report.
This inspection constituted one MSPI heat removal system sample as defined in
IP 71151-05.
21 Enclosure
b. Findings
No findings of significance were identified.
.2 Mitigating Systems Performance Index - Residual Heat Removal System
a. Inspection Scope
The inspectors sampled licensee submittals for the MSPI - Residual Heat Removal
System performance indicator for the period from the third quarter 2009 through the
second quarter 2010. To determine the accuracy of the PI data reported during those
periods, PI definitions and guidance contained in the Nuclear Energy Institute (NEI)
Document 99-02, Regulatory Assessment Performance Indicator Guideline,
Revision 6, dated October 2009, were used. The inspectors reviewed the licensees
operator narrative logs, issue reports, MSPI derivation reports, event reports and NRC
Integrated Inspection Reports for the period of the third quarter 2009 through the second
quarter 2010 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 issue report
database to determine if any problems had been identified with the PI data collected or
transmitted for this indicator and none were identified. Documents reviewed are listed in
the Attachment to this report.
This inspection constituted one MSPI residual heat removal system sample as defined in
IP 71151-05.
b. Findings
No findings of significance were identified.
.2 Mitigating Systems Performance Index - Cooling Water Systems
a. Inspection Scope
The inspectors sampled licensee submittals for the MSPI - Cooling Water Systems
performance indicator for the period from the third quarter 2009 through the second
quarter 2010. To determine the accuracy of the PI data reported during those periods,
PI definitions and guidance contained in the NEI Document 99-02, Regulatory
Assessment Performance Indicator Guideline, Revision 6, dated October 2009, were
used. The inspectors reviewed the licensees operator narrative logs, issue reports,
MSPI derivation reports, event reports and NRC Integrated Inspection Reports for the
period of the third quarter 2009 through the second quarter 2010 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 issue report database to
determine if any problems had been identified with the PI data collected or transmitted
for this indicator and none were identified. Documents reviewed are listed in the
Attachment to this report.
22 Enclosure
This inspection constituted one MSPI cooling water system sample as defined in
IP 71151-05.
b. Findings
No findings of significance were identified.
4OA2 Problem Identification and Resolution (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 CAP
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 CR 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.
23 Enclosure
b. Findings
No findings were identified.
.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 6-month period from January 2010 through June 2010,
although some examples expanded beyond those dates where the scope of the trend
warranted.
The reviews also included issues documented outside of 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 review constituted a single semi-annual trend inspection sample as defined in
IP 71152-05.
b. Findings
No findings were identified.
.4 In-depth Review- Technical Specifications Compliance
a. Inspection Scope
The inspectors performed an annual follow-up of selected issues sample of the
licensees process for performing and documenting TS compliance. The inspectors
reviewed documentation in the licensees CAP, official narrative operating logs and LCO
tracking module, for compliance with site-specific administrative, operational, and
licensing procedures specifically to assess for proper control and documentation of the
entry and exit of LCO Conditions and Required Actions. Documents reviewed are listed
in the Attachment to this report.
This review constituted one in-depth problem identification and resolution sample as
defined in IP 71152-05.
b. Findings
Introduction: The inspectors identified a finding of very low safety significance (Green)
and an associated NCV for the licensees failure to follow the requirements of TS
LCO 3.0.2 by not entering TS LCO 3.3.5.1 Condition A and TS 3.3.6.1 Condition A when
24 Enclosure
reactor vessel level instruments 1B21N0673C and 1B21N0674C were declared
inoperable. Technical Specification LCO 3.0.2 requires that Upon discovery of a failure
to meet an LCO, the Required Actions of the associated Conditions shall be met, except
as provided in LCO 3.0.5 and LCO 3.0.6.
Description: On August 9, 2010, during a review of operator narrative logs, the
inspectors noted a log entry that identified the use of a TS Surveillance Requirement
(SR) Note to support the performance of WO #200322765, PDP - New PM Replace
Rosemount STU Card. This WO included a step to acquire as-found data of the card
being replaced prior to its removal. The method of acquiring this as-found data included
performing portions of surveillance test procedure SVI-B21-T0187C, ECCS/HPCS RPV
Water Level 2 and Level 8 Channel C Functional for 1B21-N673C. Additionally, this WO
step stated, Sign Off/Close Surveillance Instruction as No Credit. Surveillance test
SVI-B21-T0187C renders Reactor Vessel Level instruments 1B21N0673C and
1B21N0674C inoperable. This surveillance references the TS Surveillance Notes
associated with SR 3.3.5.1 and 3.3.6.1. These SR Notes state, in part, When a channel
is placed in an inoperable status solely for performance of required Surveillances, entry
into associated Conditions and Required Actions may be delayed. The licensee utilized
the SR Note during the performance of as-found checks using the surveillance and did
not enter the Conditions and Required Actions for the 22 minutes it took to perform the
test.
The inspectors reviewed the licensees use of the surveillance notes and determined
that the delay in entering the Conditions and Required Actions was inappropriate
because the surveillance was being performed to satisfy WO requirements, not
TS-required SRs. As a result, the licensee declared the instruments inoperable but
did not enter the Conditions or Required Actions for the associated LCOs. This is
contrary to the requirements of TS LCO 3.0.2 which states Upon discovery of a failure
to meet an LCO, the Required Actions of the associated Conditions shall be met, except
as provided in LCO 3.0.5 and LCO 3.0.6. Limiting Condition of Operation 3.0.5 and
LCO 3.0.6 did not apply in this situation.
An additional review of recent narrative log entries identified several instances of
misapplication of the same surveillance notes. The longest time period the LCO was not
adhered to was 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 47 minutes. In combination with the replacement and
subsequent operability testing, the instrument(s) were inoperable on several different
occasions, for a sum total of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> and 29 minutes. The LCO allows the instrument(s)
to be inoperable for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before any additional actions are required. The
inspectors did not identify any instances where the LCO Required Action times were
exceeded.
Analysis: The inspectors determined that the licensees failure to follow TS LCO 3.0.2
constituted a performance deficiency. Specifically, the licensee did not enter the LCOs
and Required Actions for inoperable TS equipment. The inspectors evaluated the
performance deficiency in accordance with IMC 0612, Appendix B, Issue Screening.
This performance deficiency was not similar to any of the examples in IMC 0612,
Appendix E, Examples of Minor Issues," but was characterized as more than minor
because it impacted the Equipment Performance attribute of the Mitigating Systems
Cornerstone, and adversely affected the cornerstone objective to ensure the availability,
reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences (i.e., core damage); and if left uncorrected it could lead to a
more significant safety concern.
25 Enclosure
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 3b for the Mitigating Systems
Cornerstone. The inspectors determined the finding was of very low safety significance
(Green) because it was not a design/qualification deficiency, did not represent a loss of
system safety function, did not result in a loss of function of a single train for greater than
its TS-allowable outage time, did not result in a loss of function of nonsafety-related
risk-significant equipment and was not risk significant due to external events.
This finding has a cross-cutting aspect in the decision making component of the Human
Performance cross-cutting area (per IMC 0310 H.1(b)), because the licensee did not use
conservative assumptions to ensure the proposed action was safe. Specifically, the
licensee incorrectly used the TS SR Note to satisfy maintenance requirements.
Enforcement: The inspectors determined that the finding represents a violation of
regulatory requirements because it involved improper implementation of TS. The
licensee utilized TS SR Notes while performing surveillances to satisfy maintenance
WOs. In accordance with TS LCO 3.0.2, in these cases entry into TS LCO 3.3.5.1
Condition A and 3.3.6.1 Condition A is required. Contrary to the above, the licensee did
not enter the Conditions and Required Actions immediately upon declaring TS-required
instrumentation inoperable. Because this finding is of very low safety significance and
because it was entered into the licensees CAP as CR 10-81162, this violation is being
treated as an NCV consistent with Section 2.3.2 of the NRC Enforcement Policy.
(NCV 05000440/2010004-02; Failure to Comply with Technical Specification LCOs
When Reactor Vessel Level Instruments Were Declared Inoperable.)
4OA3 Follow-up of Events and Notices of Enforcement Discretion (71153)
.1 (Closed) Licensee Event Report 05000440/2010-003: Loss of Control Rod Drive Header
Pressure Results in Manual RPS Actuation
a. Inspection Scope
On May 11, 2010, a manual actuation of RPS was inserted to comply with TS because
of multiple accumulators being inoperable coincident with the inability to restore control
rod drive (CRD) charging header pressure. A trip unit failure caused an invalid
loss-of-coolant accident (LOCA) initiation signal and resulted in the load shed of the
XH12 stub bus. Due to an abnormal electrical lineup, both CRD pumps tripped and they
were unable to be restarted. The licensee replaced the trip unit and restored the CRD
system to its normal configuration. The licensee documented the failed equipment in
CR 10-76727. The inspectors reviewed this Licensee Event Report (LER) and did not
identify any findings or violations of NRC requirements. Documents reviewed as part of
this inspection are listed in the attachment. This LER is closed.
This event follow-up review constituted one sample as defined in IP 71153-05.
26 Enclosure
.2 (Closed) Retraction of Event Notification 45815: Loss of Safety Function to Control the
Release of Radioactive Material
a. Inspection Scope
On April 6, 2010, the licensee initiated an event notification (EN) related to a loss of
safety function involving five containment isolation valves. Specifically, the licensee
reported that they had a potential loss of safety function for the ability to control the
release of radioactive material. This was due to a loss of power to the LOCA isolation
logic associated with containment penetration single valve isolations. On June 6, 2010,
the licensee retracted this notification. The licensee evaluated the condition and
determined the containment penetrations were still able to perform their design function.
The inspectors reviewed the information contained in the evaluation, and did not identify
any findings or violations related to the licensees retraction. This EN retraction is
closed.
This event follow-up review constituted one sample as defined in IP 71153-05.
4OA5 Other Activities
.1 (Closed) Unresolved Item 05000440/2010003-06: Failure to Perform a Hydrostatic Test
in Accordance with ASME Code
a. Inspection Scope
This Unresolved Item (URI) is associated with the licensees actions following a repair to
ESW underground piping in the spring of 2009. The licensee conducted only a leak test
of the repairs rather than a hydrostatic test, and the coupling used to repair the pipe leak
was not hydrostatically tested for 10 minutes prior to installation in the system. After
further review of the repair process and interaction with the ASME code committee, the
inspectors determined that the Dresser coupling used to repair the pipe did not meet the
ASME code definition of a component, and was therefore not required to be
hydrostatically tested. This URI is closed and no further actions are required.
.2 Institute of Nuclear Power Operations Plant Assessment Report Review
a. Inspection Scope
The inspectors reviewed the final report for the Institute of Nuclear Power Operations
(INPO) plant assessment of Perry station conducted in August 2009. The inspectors
reviewed the report to ensure that issues identified were consistent with the NRC
perspectives of licensee performance and to verify whether any significant safety issues
were identified that required further NRC follow-up.
b. Findings
No findings of significance were identified.
27 Enclosure
4OA6 Meetings
.1 Exit Meeting
The inspectors presented the inspection results to the Site Vice-President,
Mr. Mark Bezilla, and other members of licensee management on October 6, 2010.
The inspectors asked the licensee whether any materials examined during the
inspection should be considered proprietary. No proprietary information was identified.
.2 Interim Exit Meetings
An interim exit meeting was conducted for radiological hazard assessment and exposure
controls, in-plant airborne radioactivity control and mitigation, and occupational dose
assessment with Mr. T. Jardine and other members of the Perry staff on July 16, 2010.
The inspectors confirmed that none of the potential report input discussed was
considered proprietary.
4OA7 Licensee-Identified Violations
The following violation of very low safety significance (Green) was identified by the
licensee and is a violation of NRC requirements which meets the criteria of the NRC
Enforcement Policy, for being dispositioned as an NCV.
- On August 25, 2010, the licensee identified a failure to meet the requirements of TS 5.5.9, Diesel Fuel Oil Testing Program requirements by failing to conduct the test for
viscosity at the prescribed temperature when receiving new fuel oil. The cause was
a failure to make appropriate procedure changes when the site implemented a
license change request that revised this TS requirement. Specifically, in September
1990, when the license change request was implemented by the site, the
temperature specified in SR 3.8.3.3 changed from 100 °F to 40 °C. Following this
change, the site did not recognize that the fuel oil viscosity test procedures
containing the prescribed testing temperature needed to be changed to align with the
new TS requirements, and therefore, the procedures incorrectly continued to reflect
the temperature cited in the previous TS version. Licensee personnel had been
testing the fuel oil in accordance with these procedures for approximately 20 years.
Corrective actions include sampling of all three fuel storage tanks for the diesel
generators, testing the samples for viscosity at the correct temperature requirement,
and implementation of procedural changes to incorporate the revised temperature.
All other TS-required surveillances of fuel oil properties were properly performed and
completed as required to ensure current operability. The violation was determined to
be of low safety significance through a licensee evaluation of risk. The licensee
entered this performance deficiency into the CAP as CR 10-81724, Fuel Oil Samples
Not Analyzed per Tech Specs.
ATTACHMENT: SUPPLEMENTAL INFORMATION
28 Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
M. Bezilla, Vice President Nuclear
D. Evans, Work and Outage Management Director
J. Grabnar, Site Engineering Director
H. Hanson, Performance Improvement Director
T. Jardine, Operations Manager
K. Krueger, Plant General Manager
P. McNulty, Radiation Protection Manager
M. Stevens, Maintenance Director
J. Tufts, Chemistry Manager
Other
C. OClare, Ohio Department of Health
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
Opened and Closed
05000440/2010004-01 NCV Unacceptable Preconditioning of RHR Valve Prior to ASME
In-Service Testing (1R22)05000440/2010004-02 NCV Failure to Comply with Technical Specification LCOs When
Reactor Vessel Level Instruments Were Declared
Inoperable (4OA2.4)
Closed
05000440/2010003-06 URI Failure to Hydrostatically Test Replacement Components in
Accordance with ASME (Section 4OA5.1)
05000440/2010-003 LER Loss of Control Rod Drive Header Pressure Results in
Manual RPS Actuation (Section 4OA3.1)
Discussed
45815 EN Retraction of Event Notification 45815: Loss of Safety
Function to Control the Release of Radioactive Material
(Section 4OA3.2)
1 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 Adverse Weather
CR 10-80444; Security Project - North-Side Concrete T Wall Installation Issues
Drawing 743-0013-00000; Topography and Storm Drain Composite; Revision D
EER 600631290; Perform Evaluation to Determine Locations of Drainage Gaps in Installed
T-Walls; dated August 4, 2010
1R04 Equipment Alignment
CR 08-42257; Annulus Exhaust Gas Treatment System (AGETS) "A" Train Low Flow
Adjustment; dated June 20, 2008
CR 10-72614; Unplanned Fire Suppression Impairment for Annulus Exhaust Gas Treatment
System; dated March 4, 2010
CR 08-34483; Annulus Exhaust Gas Treatment System Flow Indication Low Flow; dated
January 29, 2008
CR 07-31871; AEGTS B Discharge Damper Is Not Functioning Correctly; dated
December 21, 2007
Drawing 912-0605-00000; Reactor Building Annulus Exhaust Gas Treatment; Revision W
PYBP-POS-2-2; Annulus Exhaust Gas Treatment System A (B) Outage Protected Equipment
Posting Checklist; Revision 10
PNPP No. 10392; Annulus Exhaust Gas Treatment System A (B) Outage Protected Equipment
Posting Checklist; dated July 14, 2009
SOI-M15; AEGTS System; Revision 8
VLI-M23/24; MCC, Switchgear and Miscellaneous Electrical Equipment Area HVAC System;
Revision 7
CR 10-82114; 0M23C0002B Did Not Trip with a B Train Trip Signal Present; dated
September 2, 2010
CR 10-82118; Replacement Solenoid Valve Mount Screw Holes Are Not Threaded; dated
August 31, 2010
Drawing 912-0609-00000; MCC Switchgear and Misc Electrical Equipment Areas HVAC
System and Battery Room Exhaust; Revision AA
Perry Plant Health Report 2010-2 for P42 - Emergency Closed Cooling System
SOI-P42; Emergency Closed Cooling System; Revision 16
VLI-P42; Emergency Closed Cooling System; Revision 15
Drawing 302-0621-00000; Emergency Closed Cooling System; Revision SS
Drawing 208-0041-00002; Reactor Protection System MG Set S001B
Drawing 208-0041-00001; Reactor Protection System MG Set S001A
CR 10-81707; Overheating on Voltage Regulator for RPS MG Set B; dated August 25, 2010
2 Attachment
1R05 Fire Protection (Annual/Quarterly)
PAP-1910; Fire Protection Program; Revision 19
P54-24; Calculation of Combustible Loading and Allowable Limits for Fire Loading; Revision 4
FPI-0IB; Pre-Fire Plan Instruction - Intermediate Building; Revision 5
FPI-0CC; Pre-Fire Plan Instruction - Control Complex; Revision 8
CR 10-80981; Documentation of NRC Questions; dated August 9, 2010
CR 10-81985; Response to Questions from the NRC Resident Inspector; dated August 27, 2010
FPI-1AB; Pre-Fire Plan Instruction - Auxiliary Building; Revision 3
CR 10-82504; NRC Question Regarding Pen Seals in AX 620 West; dated September 10, 2010
1R06 Internal Flooding
PAP-0204; Housekeeping/Cleanliness Control Program; Revision 24
ARI-H13-P601-0018; Leak Detection; Revision 13
NOP-OP-1012; Material Readiness and Housekeeping Inspection Program; Revision 5
CR 10-77685; Various Through Wall Piping Leaks on N71; dated June 3, 2010
Drawing 911-0617; Auxiliary Building Drains; Revision F
1R11 Licensed Operator Requalification Program
PYBP-PTS-0005; Operator Continuing Training Program Administration; Revision 25
PYBP-POS-0027; Operator Actions from Memory; Revision 0, dated December 3, 2008
Simulator Exercise Guide OTLC-3058201010_PY_SGC1; Cycle 10 2010 Evaluated Scenario
C1; Revision 0
CR 10-80980; Unsat Training Observation - Ops Performance Improvement Time Not Properly
Used; dated August 9, 2010
CR 10-81725; Unqualified Individuals Signing as Training Coordinators; dated August 25, 2010
1R12 Maintenance Effectiveness
WO 200284303; Chg Oil Fltrs Combustion Gas Purge Unit; dated July 21, 2010
CR 10-79817; Wrong Oil Added to CGMC Reservoir; dated July 17, 2010
CR 10-80089; NRC-ID. No FME High Risk Brief Sheet in Work Order; dated July 22, 2010
CR 10-80169; Failed PMT for CGMC B Aux Oil Pump; dated July 24, 2010
Clearance EPY-M25-0005; Control Room HVAC Supply Plenum; dated September 1, 2010
LCOTR# A10-M25-032; M25/26 Inoperable, Period 5 Week 10; dated August 30, 2010
CR 10-81952; Relay Contacts do not Change State; dated August 30, 2010
CR 10-81957; Loose Fittings on Low Flow Switch; dated August 30, 2010
Drawing 912-0610-00000; Control Room HVAC and Emergency Recirculation System;
Revision FF
CR 10-82639; Maintenance HPCS Work Start Deficiencies; dated September 13, 2010
CR 10-82715; Inadequate Order for Div 3 Fuel Oil Day Tank Work; dated September 16, 2010
CR 10-82864; Grease Fitting Damaged during Disassembly; dated September 19, 2010
CR 10-82970; Less Than Adequate Contingency Planning for Div 3 DG Inspections; dated
September 21, 2010
CR 10-82989; FME Concerns Identified in Div 3 DG Room; dated September 20, 2010
CR 10-83194; PMT Could Not Be Worked as Written; dated September 24, 2010
WO 200430281; Rebuild Ball Valves to Small and Large Seals
CR 10-83134; Lower Airlock Door Air Supply Flex Hoses Possibly Defective
CR 10-82842; Lower Airlock Pneumatic System Pressure Drop Test Failed
3 Attachment
CR 10-76252; Lower Containment Airlock Reactor Door
CR 09-69338; Upper Containment Airlock Reactor Door
1R13 Maintenance Risk Assessments and Emergent Work Control
NOP-OP-1007; Risk Management; Revision 7
CR 10-80396; Perry Not Notified of Conservative Grid Ops; dated July 28, 2010
CR 10-81724; Finding - Fuel Oil Samples not Analyzed per Tech Specs; dated August 25, 2010
CR 10-81727; Diesel Fuel Oil Sample Analysis Completion Dates Inconsistent; dated August
25, 2010
CR 10-81733; Procedure Steps Signed as Performed Inappropriately; dated August 25, 2010
CR 10-82658; Water/Steam Leak From 1N27F505D (RFBP D Discharge Check Valve); dated
September 15, 2010
WO 200430709; Wire Wrap/Inject Inspection Flange; dated September 17, 2010
WO 200430710; Remove Insulation @ Valve; dated September 16, 2010
ECP 10-0570-000; Leak Sealant Device on Reactor Feedwater Booster Pump D Discharge
Check Valve (1N27F0505D); Revision 0
ECP 10-0570-001; Install and Inject Leak Sealant Device on Reactor Feedwater Booster Pump
D Discharge Check Valve (1N27F0505D); Revision 1
CR 10-82682; Div 3 DG Generator Inter Pole Side Plate Movement; dated September 15, 2010
CR 10-82992; Div 3 Diesel Generator - Migrating Exciter Field Core Plates; dated
September 22, 2010
WO 200430766; Remove Generator Rotor, Inspect for Loose Wedge Studs; dated
September 15, 2010
1R15 Operability Evaluations
CR 10-78672; 1M43 Agastat Relay Qualification Issue; dated June 22, 2010
CR 10-81023; M52 Turning Vanes Degraded; dated August 10, 2010
Prompt Functionality Assessment for Degraded TSC Ventilation Supply Fan Turning Vanes;
dated August 13, 2010
Prompt Operability Determination for Diesel Generator Building Ventilation Systems; dated
July 15, 2010
CR 10-81973; No Insulation Inside Plenum; dated August 30, 2010
eSOMS Narrative Logs dated September 2, 2010
Prompt Operability Determination for ECC to FPCC Heat Exchanger Bypass Valve Stroke Time
Testing Failure; August 24, 2010
CR 10-81623; OP42F0255B Failed Stroke Closed Test; dated August 23, 2010
1R18 Permanent/Temporary Modifications
Perry Plant Health Report 2010-2 for Temporary Modifications
NOP-CC-2003; Engineering Changes; Revision 14
NORM-CC-2001; Engineering Change Process Flowcharts; Revision 00
ECP 10-0020-0000; Reference Documents - Hot Surge Tank Low Level Alarm from Level
Transmitter Signal; Revision 0
ECP 10-0020-0001; Hot Surge Tank Low Level Alarm from Level Transmitter Signal; Revision 3
WO 200399695; Hot Surge Tank Low Level Alarm; dated May 15, 2010
NOBP-ER-3003-01; Temporary Modification Review Checklist; Revision 00
CR 09-67788; Host Surge Tank (HST) Level Low Alarm Locked In; dated November 15, 2009
Drawing 302-0081-00000; Feedwater; Revision BBB
4 Attachment
Drawing 302-0101-00000; Condensate System; Revision TT
Drawing 208-0149-00002; MDFP Auto Start Logic & RFBP Auto Start Logic; Revision S
CR 10-82802; Potential Single Failure Vulnerability with Hot Surge Tank Temp Mod; dated
September 16, 2010
1R19 Post-Maintenance Testing
SVI-B21-T0137F; ECCS Drywell Pressure High Channel F Functional for 1B21-N694F;
Revision 5
PTI-M23-P0005; Emergency Service Water Pump House Ventilation System Train B Damper
Stroking; Revision 5
WO 200323496; Replace Rosemount MTU Card; dated August 11, 2010
WO 200323644; Replace Keylock Control Switch 1B21C-S27A; dated August 4, 2010
WO 200340398; Replace and Perform Calibration Check of 1M15D0001B Instrumentation;
dated August 25, 2010
WO 200327715; Replace AEGT Fan B Motor; dated August 25, 2010
WO 200290571; Replace SLS/MTR/Oil Hydramotor at ESW B Outlet Damper; dated
September 6, 2010
WO 200333304; MERP - Replace Utility Station w/NUS; dated September 6, 2010
CR 10-81632; Temperature Switch Found Tripped; dated August 23, 2010
CR 10-81633; RFACR: Damaged Field Conductor to Motor; dated August 23, 2010
WO 200328863; Replace Cntrl Relays in EH1304 Cubicle; dated September 20, 2010
SOI-R22; Metal Clad Switchgear 5-15 KV; Revision 25
CR 10-82852; Unexpected Reading Obtained during Functional Testing; dated September 19,
2010
SVI-E22-T1319; Diesel Generator Start and Load Division 3; Revision 15
CR 10-83148; Div 3 Emergency Diesel Generator Failure to Start During Testing; dated
September 24, 2010
CR 10-83163; Generator Stator Temperature Monitor is Erratic and Unreliable; dated
September 24, 2010
CR 10-83181; Div 3 DG Additional Tagging Points Requested; dated September 24, 2010
1R22 Surveillance Testing
SVI-E12-T2001; RHR A Pump and Valve Operability Test; Revision 26
SVI-E12-T1194; LPCI Pump A Discharge Low Flow (Bypass) Channel Functional for 1E12-
N652A; Revision 8
SVI-E51-T2001; RCIC Pump and Valve Operability Test; Revision 32
CR 01-79624; NRC-Identified Concern for Pre-conditioning Valve During Surveillance Testing;
dated July 12, 2010
NOP-ER-3204; Inservice Testing Program; Revision 1
eSOMS Narrative Logs dated July 7-8, 2010
SVI-P45-T2002; ESW Pump B and Valve Operability Test; Revision 26
SVI-R10-T5227; Off-Site Power Availability Verification; Revision 2
5 Attachment
2RS1 Radiological Hazard Assessment and Exposure Controls
CR 09-56065; Containment Vessel Drywell Purge Degraded Flows Impacting Refuel Floor;
dated March 25, 2009
CR 09-57294; Boundary Exceeded Radiological Controlled Area (RCA); dated April 16, 2009
CR 09-60436; Dose Rates in the P5480405 Condensate Backwash Receiving Tanl Higher than
Expected; dated June 11, 2009
CR 09-62628; Radioactive Material Found Outside the RCA; dated August 2, 2009
CR 09-63398 and Associated Apparent Cause Evaluation; Platform Found Outside with Fixed
Contamination; dated August 18, 2009
CR 09-66069; RISB Radioactive Material Inventory Discrepancies; dated October 16, 2009
CR 10-76774; Radiological Issues Associated with Division 2 ECC LOCA Initiation; dated
May 11, 2010
CR-09-54403; RFO-12 Elevated Airborne Levels During Separator Lift; dated February 28, 2009
HPI-C0014; Radlock key Issue; Revision 01
HPI-H0004; Identification of Radioactive Materials and Release of Materials from RCAs;
Revision 22
HPI-K0009; Operation of the WARF, RISB and OSSC Yard; Revision 0
HPI-L0004; Source Control Documentation and Inventory; Revision 8
NOPB-NF-3102; Control of Non-Special Nuclear Material in the Fuel Pools; Revision 00
NOP-OP-4101; Access Controls for Radiologically Controlled Areas; Revision 01
NOP-OP-4102; Radiological Postings, Labeling, and Markings; Revision 05
NOP-OP-4107; Radiation Work Permit; Revisions 4 and 5
NRC Form 748; National Source Tracking Transaction Report; dated January 12, 2009
NSTS Annual Inventory Reconciliation; dated September 9, 2009, and January 29, 2010
PNPP No. 10280; Sealed Source Leak Test Data Sheet HPI-L0004; dated January 13, 2010
PNPP No. 7445; Sealed Source Leak Test Data Sheet ORM 6.4.2; dated January 13, 2010
SVI-E31-T5190; Sealed Source Leak Test and Inventory; Revision 5
TEDE ALARA Evaluations for ALARA Plan Nos. 09-6018-02, 09-6041-00 and
10-0066; dates October 2008 and February 2010
2RS3 In-Plant Airborne Radioactivity Control and Mitigation
Air Sample Records/Collection and Evaluation Forms for Various Work Activities and Locations;
Various Dates in March and April 2009
CR 09-57025; Air Sampling Equipment Found with Expired Calibration; dated April 09, 2009
EP-Emergency Plan for Perry Nuclear Power Plant Docket Nos. 50-440; Revision 30
HPI-G0007; Maintenance of Respiratory Protective Equipment and Operation of the Respirator
Cleaning / Issue Facilities; Revision 21
HPI-G0008; Requalification of Respirators; Revision 07
HPI-L0003; Equipment History; Revision 06
HRI-0003; Respirator Qualification Health Assessment; Revision 02
NOP-OP-4301; Respiratory Protection Program; Revision 01
NOP-OP-4302; Issuing Respiratory Protection; Revision 00
NOP-OP-4303; Respirator Quantitative Fit Test Portacount PRO 8030; Revision 01
NOP-OP-4310; Firehawk M7 Self Contained Breathing Apparatus; Revision 04
NOP-OP-4702; Air Sampling; Revision 01
PSI-0022; Emergency Plan Training program; Revision 03
PYBP-RPS-0038; Radiologically Controlled Area HEPA Ventilation and HEPA Vacuum Unit
Program; Revision 01
6 Attachment
2RS4 Occupational Dose Assessment
ALARA Plan 09-6040; Suppression Pool Cleaning and Inspection; Revision 03
NOP-OP-4204; Special External Exposure Monitoring; Revision 03
NOP-OP-4204-04; Effective Dose Equivalent Dose Determination; Revision 01
NOP-OP-4205; Dose Assessment; Revision 03
NOP-OP-4206; Bioassay Program; Revision 00
NOP-OP-4503; Personnel Contamination Monitoring; Revision 02
Radiological Engineering Assessment; Source term Determination for Cycle 12 Operations;
Undated
RWP 09-6040; RFO-12 Suppression Pool Diving Activities; Revision 03
4OA1 Performance Indicator Verification
NOBP-LP-4012; NRC Performance Indicators; Revision 3
NOBP-LP-4012-06; MSPI Data Sheets for Heat Removal System from July 2009 to June 2010;
Revision 2
NOBP-LP-4012-07; MSPI Data Sheets for Residual Heat Removal System from July 2009 to
June 2010; Revision 2
NOBP-LP-4012-19; MSPI Data Sheets for Emergency Service Water from July 2009 to
June 2010; Revision 2
Mitigating Systems Performance Index Basis Document; Revision 4
PYBP-DES-0011; Mitigating Systems Performance Index; Revision 1
eSOMS Narrative Logs; July 2009 to June 2010
List of CRs for all MSPI monitored systems; July 2009 to June 2010
MSPI Derivation Reports for all MSPI monitored systems; June 2010
4OA2 Identification and Resolution of Problems
CRs for the period January 1, 2010, through June 30, 2010
CR 10-81162; Potential Misapplication of TS Note; dated August 12, 2010
eSOMS Narrative Logs; July 2010, to August 2010
eSOMS Action Tracking; July 2010 to August 2010
WO 200322765; PDP - New PM Replace Rosemount STU Card; dated August 9, 2010
SVI-B21-T0187C, ECCS/HPCS RPV Water Level 2 and Level 8 Channel C Functional for
1B21-N673C; Revision 6
4OA3 Follow-up of Events and Notices of Enforcement Discretion
LER 2010-003; Loss of Control Rod Drive Header Pressure Result in Manual RPS Actuation;
dated July 12, 2010
CR 10-74904; During SVI-E12T0146 Performance, Operations Received Unexpected
Annunciators; dated April 4, 2010
4OA7 Licensee-Identified Findings
CR 10-81724; Fuel Oil Samples not Analyzed per Tech Specs; dated August 25, 2010
7 Attachment
LIST OF ACRONYMS USED
AEGTS annulus exhaust gas treatment system
ALARA as-low-as-reasonably-achievable
ASME American Society of Mechanical Engineers
CAP corrective action program
CFR Code of Federal Regulations
CR condition report
ECC emergency closed cooling
EDG emergency diesel generator
ESW emergency service water
FENOC FirstEnergy Nuclear Operating Company
HEPA high-efficiency particulate air
HP health physics
IMC Inspection Manual Chapter
IP Inspection Procedure
IR Inspection Report
IST inservice testing
LCO limiting condition for operation
LER Licensee Event Report
LPCI low pressure core injection
MSPI mitigating systems performance index
NCV non-cited violation
NEI Nuclear Energy Institute
NIOSH/MSSHA National Institute for Occupational Safety and Health/
Mine Safety and Health Administration
NOP Nuclear Operating Procedure
NRC Nuclear Regulatory Commission
ONI Off-Normal Instruction
PI performance indicator
PM post-maintenance
RCIC reactor core isolation cooling
RP radiation protection
RWP radiation work permit
SCBA self-contained breathing apparatus
SDP Significance Determination Process
SR surveillance requirement
SSC structure, system, or component
SVI Surveillance Instruction
TS Technical Specification
USAR Updated Safety Analysis Report
VHRA very high radiation area
WO work order
8 Attachment
M. Bezilla -2-
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its
enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records (PARS) component of NRC's document system
(ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html
(the Public Electronic Reading Room).
Sincerely,
/RA/
Jamnes L. Cameron, Chief
Branch 6
Division of Reactor Projects
Docket No. 50-440
License No. NPF-58
Enclosure: Inspection Report 05000440/2010004
w/Attachment: Supplemental Information
cc w/encl: Distribution via ListServ
DISTRIBUTION:
See next page
DOCUMENT NAME: G:\DRPIII\Perry\PER 2010004.docm
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 RIII
NAME PVoss:dtp JCameron
DATE 10/28/10 10/29/10
OFFICIAL RECORD COPY
Letter to M. Bezilla from J. Cameron dated October 29, 2010
SUBJECT: PERRY NUCLEAR POWER PLANT NRC INTEGRATED
INSPECTION REPORT 05000440/2010004
DISTRIBUTION:
RidsNrrPMPerry
RidsNrrDorlLpI3-2
RidsNrrDirsIrib Resource
DRPIII
DRSIII
Patricia Buckley
ROPreports Resource