ML081350692
| ML081350692 | |
| Person / Time | |
|---|---|
| Site: | Kewaunee  |
| Issue date: | 05/14/2008 |
| From: | Michael Kunowski NRC/RGN-III/DRP/B5 |
| To: | Christian D Virginia Electric & Power Co (VEPCO) |
| References | |
| IR-08-002 | |
| Download: ML081350692 (48) | |
See also: IR 05000305/2008002
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION III
2443 WARRENVILLE ROAD, SUITE 210
LISLE, IL 60532-4352
May 14, 2008
Mr. David A. Christian
President and Chief Nuclear Officer
Virginia Electric and Power Company
Innsbrook Technical Center
5000 Dominion Boulevard
Glen Allen, VA 23060-6711
SUBJECT:
KEWAUNEE POWER STATION - NRC INTEGRATED
INSPECTION REPORT 05000305/2008002
Dear Mr. Christian:
On March 31, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated
inspection at your Kewaunee Power Station. The enclosed report documents the inspection
findings, which were discussed on April 9, 2008, with Mr. Steve Scace and other members of
your staff.
The inspection examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
Based on the results of this inspection, one NRC-identified and one self-revealed finding of very
low safety significance were identified. The findings involved a violation of NRC requirements.
However, because of their very low safety significance, and because the issues were entered
into your corrective action program, the NRC is treating the issues as Non-Cited Violations
(NCVs) in accordance with Section VI.A.1 of the NRC Enforcement Policy.
If you contest the subject or severity of an NCV, you should provide a response within
30 days of the date of this inspection report, with the basis for your denial, to the
U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC
20555-0001, with 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 Kewaunee Power Station.
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
Mr. D. Christian
-2-
(ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html
(the Public Electronic Reading Room).
Sincerely,
/RA/
Michael Kunowski, Chief
Branch 5
Division of Reactor Projects
Docket No. 50-305
License No. DPR-43
Enclosure:
Inspection Report 05000305/2008002
w/Attachment: Supplemental Information
cc w/encl:
S. Scace, Site Vice President
T. Webb, Director, Nuclear Safety and
Licensing
C. Funderburk, Director, Nuclear Licensing
and Operations Support
T. Breene, Manager, Nuclear Licensing
L. Cuoco, Esq., Senior Counsel
D. Zellner, Chairman, Town of Carlton
J. Kitsembel, Public Service Commission of Wisconsin
P. Schmidt, State Liaison Officer, State of Wisconsin
Mr. D. Christian
-2-
(ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html
(the Public Electronic Reading Room).
Sincerely,
/RA/
Michael Kunowski, Chief
Branch 5
Division of Reactor Projects
Docket No. 50-305
License No. DPR-43
Enclosure:
Inspection Report 05000305/2008002
w/Attachment: Supplemental Information
cc w/encl:
S. Scace, Site Vice President
T. Webb, Director, Nuclear Safety and
Licensing
C. Funderburk, Director, Nuclear Licensing
and Operations Support
T. Breene, Manager, Nuclear Licensing
L. Cuoco, Esq., Senior Counsel
D. Zellner, Chairman, Town of Carlton
J. Kitsembel, Public Service Commission of Wisconsin
P. Schmidt, State Liaison Officer, State of Wisconsin
DOCUMENT NAME: G:\\KEWA\\KEW 2008 002.doc
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
RIII
NAME
KBarclay
MKunowski
for SBurton
MKunowski
DATE
5/12/08
5/14/08
5/14/08
OFFICIAL RECORD COPY
Letter to D. Christian from M. Kunowski dated May 14, 2008
SUBJECT:
KEWAUNEE POWER STATION NRC INTEGRATED INSPECTION REPORT
DISTRIBUTION:
DXC1
TEB
RidsNrrDirsIrib
MAS
KGO
JKH3
Kewaunee SRI
CAA1
LSL (electronic IRs only)
C. Pederson, DRP (hard copy - IRs only)
DRPIII
DRSIII
PLB1
TXN
ROPreports@nrc.gov (inspection reports, final SDP letters, any letter with an IR number)
Enclosure
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No:
50-305
License No:
Report No:
Licensee:
Dominion Energy Kewaunee, Inc.
Facility:
Kewaunee Power Station
Location:
Kewaunee, WI
Dates:
January 1, 2008, through March 31, 2008
Inspectors:
S. Burton, Senior Resident Inspector
P. Higgins, Resident Inspector
J. Cassidy, Health Physicist
K. Barclay, Reactor Engineer
R. Langstaff, Senior Reactor Inspector
Approved by:
M. Kunowski, Chief
Branch 5
Division of Reactor Projects
Enclosure
TABLE OF CONTENTS
SUMMARY OF FINDINGS .........................................................................................................1
REPORT DETAILS.....................................................................................................................3
Summary of Plant Status.........................................................................................................3
1.
REACTOR SAFETY.....................................................................................................3
1R01
Adverse Weather Protection (71111.01) ............................................................3
1R04
Equipment Alignment (71111.04).......................................................................4
1R05
Fire Protection (71111.05) .................................................................................8
1R11
Licensed Operator Requalification Program (71111.11).....................................9
1R12
Maintenance Effectiveness (71111.12) ..............................................................9
1R13
Maintenance Risk Assessments and Emergent Work Control (71111.13)........10
1R15
Operability Evaluations (71111.15) ..................................................................11
1R18
Plant Modifications (71111.18).........................................................................12
1R19
Post-Maintenance (PM) Testing (71111.19).....................................................12
1R20
Outage Activities (71111.20)............................................................................15
1R22
Surveillance Testing (71111.22).......................................................................16
1EP6
Drill Evaluation (71114.06)...............................................................................19
2.
RADIATION SAFETY.................................................................................................20
2OS1
Access Control to Radiologically Significant Areas (71121.01) ........................20
4.
OTHER ACTIVITIES ..................................................................................................23
4OA2
Identification and Resolution of Problems (71152)...........................................23
4OA3
Follow-up of Events and Notices of Enforcement Discretion (71153)...............24
4OA5
Other Activities.................................................................................................25
4OA6
Management Meetings ....................................................................................27
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 ................................................................................................15
1
Enclosure
SUMMARY OF FINDINGS
IR 05000305/2008002; 01/01/2008 - 03/31/2008; Kewaunee Power Station; Equipment
Alignment and Post-Maintenance Testing.
This report covers a three-month period of inspection by resident inspectors and announced
baseline inspections by regional inspectors. Two Green findings, one NRC-identified and one
self-revealed, were identified by the inspectors. These findings were considered Non-Cited
Violations (NCVs) of NRC regulations. The significance of most findings is indicated by their
color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance
Determination Process (SDP). Findings for which the SDP does not apply may be Green or be
assigned a severity level after NRC management review. The NRCs program for overseeing
the safe operation of commercial nuclear power reactors is described in NUREG-1649,
Reactor Oversight Process, Revision 4, dated December 2006.
A.
NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
Green. A finding of very low safety significance (Green) and an associated NCV
of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings,
was identified by the inspectors for the licensees failure to install scaffolding in
accordance with station procedures. Specifically, more than ten examples where
scaffolding was built within 2-inches of safety-related systems without an engineering
evaluation, and six examples where non-seismic scaffolding was built in safety-related
areas were identified. The licensee suspended all scaffold building pending the
completion of their corrective actions. The corrective actions included training scaffold
builders on proper scaffold building techniques and how to identify operational and
seismic concerns, revising procedures for scaffold building to address operations and
engineering involvement in the scaffold building process, and a complete plant
walkdown of all scaffolding by engineering or operations.
This finding was more than minor because it was associated with the equipment
performance attribute of the Mitigating Systems cornerstone and affected the
cornerstone objective to ensure the availability, reliability, and capability of systems that
respond to initiating events to prevent undesirable consequences. Specifically, the
improperly installed scaffolding could have impeded or prevented proper operation of the
safety-related components. Using Attachment 4 of IMC 0609, the inspectors answered
no to all the screening questions in the SDP Phase 1 Screening Worksheet in the
Mitigating Systems column; therefore, this finding is of very low safety significance
(Green). The inspectors determined that this finding had a cross-cutting aspect in the
area of problem identification and resolution, corrective action program, because the
licensee did not take appropriate corrective actions to address safety issues and
adverse trends in a timely manner. (P.1(d)) (Section 1R04.1)
Cornerstone: Barrier Integrity
Green. A finding of very low safety significance (Green) and an associated NCV of
10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was
identified by the inspectors following surveillance testing of containment isolation valve
LOCA-3A in accordance with plant procedure SP 55 167 4B, "Post LOCA Valves Timing
2
Enclosure
Test (IST) from Local Panel-Train B." Specifically, the licensee failed to initiate a
condition report in accordance with procedure PI-KW-200, Corrective Action, following
a review of the test results by the inservice testing program engineer who subsequently
identified a potential condition which called into question the operability of LOCA-3A.
The finding was more than minor in accordance with IMC 0612, Power Reactor
Inspection Reports, Appendix B, Issue Screening, dated September 20, 2007,
because the finding was associated with the structure, system and component (SSC)
and barrier performance attribute of the Barrier Integrity Cornerstone and affected the
cornerstone objective to provide reasonable assurance that the physical design barriers
(fuel cladding, reactor coolant system, and containment) protect the public from
radionuclide releases caused by accidents or events. Specifically, the licensee failed to
implement the provisions of Corrective Action Procedure, PI-KW-200, which resulted in a
failure to ensure operability of containment isolation valve LOCA-3A. The inspectors
also determined that the primary cause for this finding was related to the cross-cutting
area of human performance, work practices, because personnel have been trained in
need for procedural use and adherence but did not follow applicable procedures.
(H.4(b)) (Section 1R19)
B.
Licensee-Identified Violations
No violations of significance were identified.
3
Enclosure
REPORT DETAILS
Summary of Plant Status
Kewaunee operated at full power during the entire first quarter of 2008 until early on
March 29, 2008, when the unit was shutdown for a scheduled refueling outage.
1.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Control, and
1R01 Adverse Weather Protection (71111.01)
.1
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 also reviewed the abnormal operating procedure for mitigating
the design basis flood to ensure it could be implemented as written.
This inspection constitutes one external flooding sample as defined in Inspection
Procedure 71111.01-05.
b.
Findings
No findings of significance were identified.
.2
Readiness For Impending Adverse Weather Condition - Extreme Cold Conditions
a.
Inspection Scope
Extreme cold conditions were forecast in the vicinity of the facility for
January 29 - 30, 2008. On these dates, the inspectors reviewed the licensees
preparation and performance for the cold weather including external equipment
walk-downs, reviews of the cold weather checklist and reviews of susceptible systems in
the auxiliary and turbine buildings because their safety-related functions could be
affected or required as a result of the extreme cold conditions forecast for the facility.
The inspectors observed insulation, heat trace circuits, space heater operation, and
weatherized enclosures to ensure operability of affected systems. The inspectors
reviewed licensee procedures and discussed potential compensatory measures with
4
Enclosure
control room personnel. The inspectors focused on plant managements actions for
implementing the stations procedures for ensuring adequate personnel for safe plant
operation and emergency response would be available. Specific documents reviewed
during this inspection are listed in the Attachment.
This inspection constitutes one readiness for impending adverse weather condition
sample as defined in Inspection Procedure 71111.01-05.
b.
Findings
No findings of significance were identified.
.3
Readiness For Impending Adverse Weather Condition - Heavy Snowfall & Ice
Conditions
a.
Inspection Scope
On February 18, 2008, a winter weather advisory was issued for expected icing and
snow squalls. The inspectors observed the licensees preparations and planning for the
significant winter weather potential. The inspectors reviewed licensee procedures and
discussed potential compensatory measures with control room personnel. The
inspectors focused on plant managements actions for implementing the stations
procedures for ensuring adequate personnel for safe plant operation and emergency
response would be available. The inspectors conducted a site walkdown including
walkdowns of various plant structures and systems to check for maintenance or other
apparent deficiencies that could affect system operations during the predicted significant
weather. The inspectors also reviewed corrective action program (CAP) items to verify
that the licensee was identifying adverse weather issues at an appropriate threshold and
entering them into their CAP in accordance with station corrective action procedures.
Specific documents reviewed during this inspection are listed in the Attachment.
This inspection constitutes one readiness for impending adverse weather condition
sample as defined in Inspection Procedure 71111.01-05.
b.
Findings
No findings of significance were identified.
1R04 Equipment Alignment (71111.04)
.1
Quarterly Partial System Walkdowns
a.
Inspection Scope
The inspectors performed partial system walkdowns of the following risk-significant
systems:
bus 6 and emergency diesel generator following bus 6 auto inhibit relay test;
auxiliary feedwater (AFW) system A following maintenance; and
safety injection train B with train A out-of-service.
5
Enclosure
The inspectors selected these systems based on their risk significance relative to the
Reactor Safety Cornerstones at the time they were inspected. The inspectors attempted
to identify any discrepancies that could impact the function of the system, and, therefore,
potentially increase risk. The inspectors reviewed applicable operating procedures,
system diagrams, the USAR, Technical Specification (TS) requirements, Administrative
TSs, outstanding work orders (WOs), condition reports, and the impact of ongoing work
activities on redundant trains of equipment in order to identify conditions that could have
rendered the systems incapable of performing their intended functions. The inspectors
also walked down accessible portions of the systems to verify system components and
support equipment were aligned correctly and operable. The inspectors examined the
material condition of the components and observed operating parameters of equipment
to verify that there were no obvious deficiencies. The inspectors also verified that the
licensee had properly identified and resolved equipment alignment problems that could
cause initiating events or impact the capability of mitigating systems or barriers and
entered them into the CAP with the appropriate significance characterization.
Documents reviewed are listed in the Attachment.
These activities constituted three partial system walkdown samples as defined in
Inspection Procedure 71111.04-05.
b.
Findings
(1) Scaffolding in Close Proximity to Multiple Safety-Related Systems Affects Operability
Introduction: A finding of very low safety significance (Green) and an associated NCV
of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings,
was identified by the inspectors for the licensees failure to install scaffolding in
accordance with station procedures. Specifically, more than ten examples were
identified where scaffolding was built within 2-inches of safety-related systems without
an engineering evaluation, and six examples where scaffolding built in a safety-related
area was not seismically qualified.
Description: On March 11, 2008, while performing a quarterly partial system walkdown
of the AFW system, the inspectors identified scaffolding that was constructed within
2-inches of the instrument sensing line for AFW flow to the 1A steam generator without
an associated engineering evaluation. Step 4.2.5 of general maintenance procedure
GMP-127, Requirements and Guidelines for Scaffold Construction and Inspection,
Revision 17, required a 2-inch clearance or approved engineering evaluation. The
inspectors examined additional scaffolding in the area and identified that the instrument
sensing line for AFW flow to 1B steam generator also had scaffolding constructed within
2-inches without an engineering evaluation. The inspectors notified the shift manager
about the two deficiencies and continued to inspect scaffolding throughout the plant.
Subsequently, engineering evaluated the scaffolding and determined that it was
adequately braced to prevent interaction with the AFW sensing lines and would not
affect the operability.
During the expanded walkdown, the inspectors identified that scaffolding built over the
safety-related steam supply line to the turbine-driven auxiliary feedwater (TDAFW) pump
was not seismically qualified. Step 4.1.23 of procedure GMP-127 requires scaffold
built-in safety-related areas to be stabilized in accordance with Section 4.2,
Safety-Related Area Scaffold Stabilization. Engineering evaluated the scaffolding and
6
Enclosure
determined that it was not seismically qualified. The licensee declared the TDAFW
pump inoperable and entered TS 3.4.b.4.A, One Train of AFW Inoperable, while they
modified the scaffolding to meet the seismic qualification standards. In total, the
licensee modified five different sets of scaffolding over or in the vicinity of the TDAFW
pump steam supply line prior to declaring the pump operable.
The licensee began inspecting scaffolding after the NRC notified them about the first
AFW sensing line issue. During the licensees inspections they identified additional
examples where non-seismic scaffolding was built in a safety-related area and where
scaffolding was within 2-inches of safety-related components without engineering
evaluations. One set of scaffolding was built in-contact with safety-related piping for two
reactor coolant sampling outboard containment isolation valve actuators, RC-413 and
RC-423, which was also not built to the seismic qualification standards of Step 4.2.3 of
procedure GMP-127. The licensee declared both valves inoperable and entered TS 3.6.b.3.A, Inoperable Containment Isolation Valve, while they disassembled the
Analysis: The inspectors determined that the installation of scaffolding too close to
safety-related components without an engineering evaluation and the installation of
non-seismic scaffolding in the area of safety-related components, was contrary to
procedural requirements, and was a performance deficiency. The finding was
determined to be more than minor because it is associated with the equipment
performance attribute of the Mitigating System Cornerstone and affected the cornerstone
objective to ensure availability, reliability and capability of systems that respond to
initiating events to preclude undesirable consequences. Specifically, the improperly
installed scaffolding could have impeded or prevented proper operation of the
safety-related components.
The inspectors determined the finding could be evaluated using the SDP in accordance
with IMC 0609, Significance Determination Process, Attachment 0609.04,
Phase 1 - Initial Screening and Characterization of Findings, Table 4a for the Mitigating
Systems Cornerstone. The inspectors answered no to all screening questions in the
Mitigating Systems Column, therefore, the finding is of very low safety significance
(Green).
The inspectors determined that this finding had a cross-cutting aspect in the area of
problem identification and resolution, corrective action program, because the licensee
did not take appropriate corrective actions to address safety issues and adverse trends
in a timely manner. Specifically, scaffolding construction within 2-inches of
safety-related components without engineering evaluations was identified by the NRC
during the last outage and documented in CAP 038722. Additionally, in December of
2007, the NRC identified that the safety-related steam supply line to the TDAFW pump
was a safety-related area and that procedure GNP-01.31.01, Plant Cleanliness and
Storage, failed to identify it as such and prevent uncontrolled storage (CAP027377).
Both examples show that the licensee had past opportunities to identify and correct the
underlying causes of the recent scaffolding problems. (P.1(d))
Enforcement: Title 10 CFR, Part 50, Appendix B, Criterion V, Instructions, Procedures,
and Drawings, states in part that, activities affecting quality, shall be prescribed by
documented instructions, procedures, or drawings, of a type appropriate to the
circumstances and shall be accomplished in accordance with these instructions,
7
Enclosure
procedures, or drawings. Kewaunee General Maintenance Procedure GMP-127
specifies in Step 5.2.5 that scaffolding shall be no closer than 2-inches from any
safety-related equipment, unless otherwise evaluated and approved by engineering.
Procedure GMP-127 also specifies in Step 4.1.23 that a scaffold built in safety-related
areas be stabilized in accordance with Section 4.2, Safety-Related Area Scaffold
Stabilization.
Contrary to the above, the licensee failed to follow procedures during the installation of
scaffolding. Specifically, on March 11, 2008, the inspectors found scaffolding
constructed within 2-inches of safety-related components without an engineering
evaluation and non-seismic scaffolding constructed in a safety-related area. The
licensee suspended all scaffold building pending the completion of their corrective
actions. The corrective actions included training scaffold builders on proper scaffold
building techniques and how to identify operational and seismic concerns, revising
procedures for scaffold building to address operations and engineering involvement in
the scaffold building process, and a plant walkdown of all scaffolding by engineering or
operations. Because this violation was of very low safety significance and it was entered
into the licensees CAP as CAP092794, CAP092776 and CAP09279, this violation is
being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy
.2
Semi-Annual Complete System Walkdown
a.
Inspection Scope
On March 13, 2008, the inspectors performed a complete system alignment inspection
of the service water 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
pressure and temperature indications, as appropriate, component labeling, component
lubrication, component and equipment cooling, hangers and supports, operability of
support systems, and to ensure that ancillary equipment or debris did not interfere with
equipment operation. A review of a sample of past and outstanding WOs was
performed to determine whether any deficiencies significantly affected the system
function. In addition, the inspectors reviewed the CAP database to ensure that system
equipment alignment problems were being identified and appropriately resolved. The
documents used for the walkdown and issue review are listed in the Attachment.
These activities constituted one complete system walkdown sample as defined in
Inspection Procedure 71111.04-05.
b.
Findings
No findings of significance were identified.
8
Enclosure
1R05 Fire Protection (71111.05)
.1
Routine Resident Inspector Tours (71111.05Q)
a.
Inspection Scope
The inspectors conducted fire protection walkdowns which were focused on availability,
accessibility, and the condition of firefighting equipment in the following risk-significant
plant areas:
Fire Zones TU-90, -91, -92, -93, 1A and 1B emergency diesel generator rooms
and associated day tank rooms;
Fire protection Impairments;
Fire Zones TU-94, SC-70A, -70B, screen house, screen house tunnel, and CO2
room;
Fire Zones TU-22, -96, turbine building basement and turbine building
mezzanine;
Fire Zones TU -95A, -95B, -95C, auxiliary feed pump area, and 480V buses
1-51, -52, -61, -62;
Fire Zones TC-100, -101, -102, technical support center;
Fire Zones AX -23B, -25, -23D, auxiliary building 606 elevation general area;
and
Fire Zone Auxiliary Building 606, north penetration room.
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 had 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.
These activities constituted eight quarterly fire protection inspection sample as defined in
Inspection Procedure 71111.05-05.
b.
Findings
No findings of significance were identified.
9
Enclosure
1R11 Licensed Operator Requalification Program (71111.11)
.1
Resident Inspector Quarterly Review (71111.11Q)
a.
Inspection Scope
On February 11, 2008, 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
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.
This inspection constitutes one quarterly licensed operator requalification program
sample as defined in Inspection Procedure 71111.11.
b.
Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness (71111.12)
.1
Routine Quarterly Evaluations (71111.12Q)
a.
Inspection Scope
The inspectors evaluated degraded performance issues involving the following
risk-significant systems:
spent fuel pump and cooling system - preps for full core offload in outage; and
containment isolation system.
The inspectors reviewed events such as where ineffective equipment maintenance has
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;
10
Enclosure
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.
This inspection constitutes two quarterly maintenance effectiveness samples as defined
in Inspection Procedure 71111.12-05.
b.
Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)
.1
Maintenance Risk Assessments and Emergent Work Control
a.
Inspection Scope
The inspectors reviewed the licensee's evaluation and management of plant risk for the
maintenance and emergent work activities affecting risk-significant and safety-related
equipment listed below to verify that the appropriate risk assessments were performed
prior to removing equipment for work:
risk assessments for work changes during the week ending January 26, 2008,
including charging pump C isolation and restoration due to work on charging
pump B ducts seal leak, and the addition of substation work;
charging pump C being returned to operation with a seal leak to allow
maintenance on charging pump A;
charging pump C isolated due to seal leak;
spent fuel pool cooling isolated for various maintenance activities;
risk assessments for work changes during the week ending March 1, 2008,
including scope change for residual heat removal (RHR) pump seal replacement,
added substation work, date change for battery room fan coil unit work; and
emergent pre-outage activities during the week ending March 29, 2008.
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
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Enclosure
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 activities constituted six samples as defined in Inspection Procedure
71111.13-05.
b.
Findings
No findings of significance were identified.
1R15 Operability Evaluations (71111.15)
.1
Operability Evaluations
a.
Inspection Scope
The inspectors reviewed the following issues:
baseline core damage frequency threshold changes for core damage frequency
and large early release frequency;
operability evaluation for the interface between condensate storage and the AFW
system;
steam generator 1B sample valve, declared inoperable and was closed and
de-energized to meet TSs;
auxiliary building fan loading was determined to be non-conservative;
emergency diesel generator power spiked abnormally during surveillance testing;
and
pressure locking of safety injection valves SI-350A, -350B.
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 also 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.
This inspection constitutes six samples as defined in Inspection Procedure 71111.15-05
b.
Findings
No findings of significance were identified.
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Enclosure
1R18 Plant Modifications (71111.18)
.1
Temporary Plant Modifications
a.
Inspection Scope
The inspectors reviewed the following temporary modification(s):
Removal of fence and steel from main transformer bay.
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.
This inspection constitutes one temporary modification sample as defined in Inspection
Procedure 71111.18-05.
b.
Findings
No findings of significance were identified.
1R19 Post-Maintenance (PM) Testing (71111.19)
.1
PM Testing
a.
Inspection Scope
The inspectors reviewed the following PM activities to verify that procedures and test
activities were adequate to ensure system operability and functional capability:
loss-of-coolant accident valve, LOCA-3A, failed PM test following overhaul;
post-maintenance test for service water valve SW-301A following replacement of
solenoid valve SV-3033;
post-maintenance test on auxiliary building basement fan coil unit D following
inspection and back-flush;
post-maintenance test following replacement of service water pump regulators
B1 & B2;
post-maintenance test following replacement of plant equipment water pump 1B;
and
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Enclosure
post-maintenance test on steam generator power-operated relief valve SD-3A
following maintenance on the related Foxborough controller.
These activities were selected based upon the SSCs ability to impact risk. The
inspectors evaluated these activities for the following (as applicable): the effect of testing
on the plant had been adequately addressed; testing was adequate for the maintenance
performed; acceptance criteria were clear and demonstrated operational readiness; test
instrumentation was appropriate; tests were performed as written in accordance with
properly reviewed and approved procedures; equipment was returned to its operational
status following testing (temporary modifications or jumpers required for test
performance were properly removed after test completion), and test documentation was
properly evaluated. The inspectors evaluated the activities against TSs, the Updated
Final Safety Analysis Report (UFSAR), 10 CFR Part 50 requirements, licensee
procedures, and various NRC generic communications to ensure that the test results
adequately ensured that the equipment met the licensing basis and design
requirements. In addition, the inspectors reviewed corrective action documents
associated with 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.
This inspection constitutes six samples as defined in Inspection Procedure 71111.19.
b.
Findings
Failure to Follow the Provisions of Corrective Action Procedure PI-KW-200 Following
Surveillance Testing of Containment Isolation Valve LOCA-3A
Introduction: A finding of very low safety significance (Green) and an NCV
of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings,
was identified by the inspectors following surveillance testing of containment isolation
valve LOCA-3A in accordance with plant procedure SP-55-167-4B, "Post LOCA Valves
Timing Test (IST) from Local Panel-Train B." Specifically, the licensee failed to initiate a
condition report in accordance with procedure PI-KW-200, Corrective Action, following
a review of the test results by the inservice testing (IST) program engineer who
subsequently identified a potential condition which called into question the operability of
LOCA-3A.
Description: On November 27, 2007, Surveillance Procedure SP-55-167-4B,
"Post-LOCA Valve is Timing Test (IST) from Local Panel-Train B," was performed on
containment isolation valve LOCA-3A. The surveillance procedure identified that the
opening time of this valve had degraded but had not exceeded the code allowable action
value. Condition Report (CR) 025595 was written to evaluate the valve stroke time and
determine if additional actions were required. This condition report concluded that since
the opening time had not exceeded the action value, LOCA-3A remained operable,
however, a corrective action was generated to evaluate the observed change in stroke
times. On November 28, the IST program engineer completed the Corrective Action
CA022013, and documented an evaluation of the change in valve stroke times. This
conclusion documented in this corrective action stated, "Since the valve is opening
slower and closing faster the most probable cause for the change in performance would
be a control air leak." A Condition Report describing this potential control air leak was
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Enclosure
not written and an operability evaluation for such a leak was not performed. Work order
KW100309607 was initiated for inspection of the valve and controller, however,
CA022013 required no additional actions. On December 13, 2007, the WO was
canceled with no action taken. On January 11, 2008, LOCA-3A was retested to validate
stroke times based on the November 27, 2007, results and the valve failed the timing
test in both the open and close directions. The licensee entered a 24-hour action
statement per plant TSs due to an inoperable containment isolation valve.
The inspectors determined that, on November 28, 2007, CA022013 identified a probable
existing condition of a control air leak which called into question the operability of
LOCA-3A. Dominion Corrective Action Procedure, PI-KW-200, required that a Condition
Report be written upon identification that such a condition may exist on safety-related
equipment. Specifically, PI-KW-200, Attachment 1, lists 50 conditions that require a
condition report. Among the conditions listed are: number 20) "Degradation, damage,
failure, malfunctioned, or loss of plant equipment."; number 26) "And an event, condition,
or situation, which on its own, is a condition potentially adverse to quality or meets the
criteria for submitting a Condition Report, even if the item will be addressed by a
separate process"; and number 31) "structures, systems, or components that enter an
alert condition (or based on their performance trend shall enter an alert condition prior to
the next schedule surveillance) in accordance with the inservice inspection or Predictive
Analysis programs." Therefore, the inspectors concluded that the licensee failed to
implement multiple provisions of PI-KW-200 which resulted in a failure to write a
condition report and subsequent failure to perform an operability evaluation on a
containment isolation valve with what was considered at the time to be a probable
control air leak.
Analysis: The inspectors determined that the licensees failure to implement the
provisions of its corrective action procedure was a performance deficiency warranting
further review. The inspectors concluded that the finding was more than minor in
accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B,
Issue Screening, dated September 20, 2007, because the finding was associated with
the SSC and barrier performance attribute of the Barrier Integrity Cornerstone and
affected the cornerstone objective to provide reasonable assurance that the physical
design barriers (fuel cladding, reactor coolant system, and containment) protect the
public from radionuclide releases caused by accidents or events. Specifically, the
licensee failed to implement the provisions of Corrective Action Procedure, PI-KW-200,
which resulted in a failure to ensure operability of containment isolation valve LOCA-3A.
The inspectors evaluated the finding using Attachment 0609.04, of IMC 0609,
Significance Determination Process, dated January 10, 2008, and answered no to all
of the questions for the Containment Barriers Cornerstone; therefore, the finding was
determined to be of very low safety significance (Green).
The inspectors also determined that the primary cause for this finding was related to the
cross-cutting area of human performance, work practices, because personnel have been
trained in need for procedural use and adherence but did not follow applicable
procedures. Specifically, procedures which required the initiation of a condition report
when a potentially discrepant condition on a containment isolation valve was identified,
which called into question valve operability, were not followed (H.4(b)).
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Enclosure
Enforcement: Title 10 CFR, Part 50, Appendix B, Criterion V, Instructions, Procedures,
and Drawings, states in part that, activities affecting quality, shall be prescribed by
documented instructions, procedures, or drawings, of a type appropriate to the
circumstances and shall be accomplished in accordance with these instructions,
procedures, or drawings. Contrary to this, the inspectors identified that the licensee
failed to implement the provisions of Procedure PI-KW-200, Corrective Action, which
resulted in a failure to ensure operability of containment isolation valve LOCA-3A. The
licensee entered this issue into its corrective action program as condition reports
CR025595, CR091329, CR028647, CR028605 and Apparent Cause Evaluations 916,
918, and 919. Corrective actions by the licensee included additional operator crew
briefs and procedure reviews and updates as appropriate. Because this violation was of
very low safety significance (Green) and was entered into the licensees corrective
action program, this violation is being treated as an NCV, consistent with Section VI.A of
the NRC Enforcement Policy (NCV 5000305/2008002-02).
1R20 Outage Activities (71111.20)
.1
Refueling Outage Activities
a.
Inspection Scope
The inspectors reviewed the Outage Safety Plan and contingency plans for the
Kewaunee Power Station refueling outage, starting on March 29, 2008, to confirm that
the licensee had appropriately considered risk, industry experience, and previous site-
specific problems in developing and implementing a plan that assured maintenance of
defense-in-depth. During the refueling outage, the inspectors observed portions of the
shutdown and cooldown processes and monitored licensee controls over the outage
activities listed below. Documents reviewed during the inspection are listed in the
Attachment.
licensee configuration management, including maintenance of defense-in-depth
commensurate with the shutdown risk assessment for key safety functions and
compliance with the applicable TSs when taking equipment out-of-service;
implementation of clearance activities and confirmation that tags were properly
hung and equipment appropriately configured to safely support the work or
testing;
controls over the status and configuration of electrical systems to ensure that
TSs and shutdown risk assessments were met, and controls over switchyard
activities;
monitoring of decay heat removal processes, systems, and components;
controls over activities that could affect reactivity; and
licensee identification and resolution of problems related to refueling outage
activities.
This inspection overlapped the inspection period and was in progress at the end of the
period. A partial refueling outage sample as defined in Inspection Procedure
71111.20-05 was documented.
b.
Findings
No findings of significance were identified.
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Enclosure
1R22 Surveillance Testing (71111.22)
.1
Routine Surveillance Testing
a.
Inspection Scope
The inspectors reviewed the test results for the following activities to determine whether
risk-significant systems and equipment were capable of performing their intended safety
function and to verify testing was conducted in accordance with applicable procedural
and TS requirements:
emergency diesel generator A monthly availability test;
engineering safeguards train A logic test;
engineering safeguards train B logic test;
emergency diesel generator B monthly availability test;
train B component cooling water pump and valve test; and
auxiliary building special ventilation zone train B monthly test.
The inspectors observed in-plant activities and reviewed procedures and associated
records to determine whether: any preconditioning occurred; effects of the testing were
adequately addressed by control room personnel or engineers prior to the
commencement of the testing; acceptance criteria were clearly stated, demonstrated
operational readiness, and were consistent with the system design basis; plant
equipment calibration was correct, accurate, and properly documented; as left setpoints
were within required ranges; the calibration frequency was 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, 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 the safety functions; and all problems
identified during the testing were appropriately documented and dispositioned in the
CAP. Documents reviewed are listed in the Attachment.
This inspection constitutes six routine surveillance testing samples as defined in
Inspection Procedure 71111.22.
b.
Findings
No findings of significance were identified.
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Enclosure
.2
Inservice Testing Surveillance
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:
post loss-of-coolant-accident valves timing test (IST) from local panel - train B.
The inspectors observed in-plant activities and reviewed procedures and associated
records to determine whether: any preconditioning occurred; effects of the testing were
adequately addressed by control room personnel or engineers prior to the
commencement of the testing; acceptance criteria were clearly stated, demonstrated
operational readiness, and were consistent with the system design basis; plant
equipment calibration was correct, accurate, and properly documented; as left setpoints
were within required ranges; and the calibration frequency were in accordance with TSs,
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 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.
This inspection constitutes one inservice inspection sample as defined in Inspection
Procedure 71111.22.
b.
Findings
No findings of significance were identified.
.3
Reactor Coolant System Leak Detection Inspection Surveillance
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
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Enclosure
function and to verify testing was conducted in accordance with applicable procedural
and TS requirements:
radiation instrument R-21 used as backup when reactor coolant system leakage
detection radiation instruments R-11 or R-12 are out-of-service.
The inspectors observed in-plant activities and reviewed procedures and associated
records to determine whether: preconditioning occurred; effects of the testing were
adequately addressed by control room personnel or engineers prior to the
commencement of the testing; acceptance criteria were clearly stated, demonstrated
operational readiness, and were consistent with the system design basis; plant
equipment calibration was correct, accurate, and properly documented; as left setpoints
were within required ranges; and the calibration frequency were in accordance with TSs,
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, 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.
This inspection constitutes one reactor coolant system leak detection inspection sample
as defined in Inspection Procedure 71111.22.
b.
Findings
No findings of significance were identified.
.4
Containment Isolation Valve Testing
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:
post loss-of-coolant accident valves - timing test train A.
The inspectors observed in-plant activities and reviewed procedures and associated
records to determine whether: any preconditioning occurred; effects of the testing were
adequately addressed by control room personnel or engineers prior to the
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Enclosure
commencement of the testing; acceptance criteria were clearly stated, demonstrated
operational readiness, and were consistent with the system design basis; plant
equipment calibration was correct, accurate, and properly documented; as left setpoints
were within required ranges; and the calibration frequency were in accordance with TSs,
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, 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.
This inspection constitutes one containment isolation valve inspection sample as defined
in Inspection Procedure 71111.22.
b.
Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP6 Drill Evaluation (71114.06)
.1
Training Observation
a.
Inspection Scope
The inspectors observed a simulator training evolution for licensed operators on
February 11, 2008, which required emergency plan implementation by a licensee
operations crew. This evolution was planned to be evaluated and included in
performance indicator data regarding drill and exercise performance. The inspectors
observed event classification and notification activities performed by the crew. The
inspectors also attended the post-evolution critique for the scenario. The focus of the
inspectors activities was to note any weaknesses and deficiencies in the crews
performance and ensure that the licensee evaluators noted the same issues and entered
them into the CAP. As part of the inspection, the inspectors reviewed the scenario
package and other documents listed in the Attachment.
This inspection constitutes one sample as defined in Inspection Procedure 71114.06-05.
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Enclosure
b.
Findings
No findings of significance were identified.
2.
RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS1 Access Control to Radiologically Significant Areas (71121.01)
.1
Review of Licensee Performance Indicators (PIs) for the Occupational Exposure
Cornerstone
a.
Inspection Scope
The inspectors reviewed the licensees occupational exposure control cornerstone PIs to
determine whether the conditions resulting in any PI occurrences had been evaluated,
and identified problems had been entered into the CAP for resolution.
This inspection represents one sample as defined in Inspection Procedure 71121.01-5.
b.
Findings
No findings of significance were identified.
.2
Plant Walkdowns and Radiation Work Permit Reviews
a.
Inspection Scope
The adequacy of the licensees internal dose assessment process for internal exposures
> 50 millirem committed effective dose equivalent was assessed.
This inspection constitutes one sample as defined in Inspection Procedure 71121.01-5.
The inspectors also reviewed the licensees physical and programmatic controls for
highly activated and/or contaminated materials (non-fuel) stored within spent fuel or
other storage pools.
This inspection represents one sample as defined in Inspection Procedure 71121.01-5.
b.
Findings
No findings of significance were identified.
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Enclosure
.3
Problem Identification and Resolution
a.
Inspection Scope
The inspectors reviewed a sample of the licensees self-assessments, audits, Licensee
Event Reports (LERs), and Special Reports related to the access control program to
determine if identified problems were entered into the CAP for resolution.
This inspection represents one sample as defined by Inspection Procedure 71121.01-5.
The inspectors reviewed corrective action reports related to access controls and high
radiation area (HRA) radiological incidents (non-PIs identified by the licensee in HRAs
<1R/hr). Staff members were interviewed and corrective action documents were
reviewed to determine whether follow-up activities were being conducted in an effective
and timely manner commensurate with their importance to safety and risk based on the
following:
Initial problem identification, characterization, and tracking;
Disposition of operability/reportability issues;
Evaluation of safety significance/risk and priority for resolution;
Identification of repetitive problems;
Identification of contributing causes;
Identification and implementation of effective corrective actions;
Resolution of NCVs tracked in the corrective action system; and
Implementation/consideration of risk-significant operational experience feedback.
This inspection constitutes one sample as defined in Inspection Procedure 71121.01-5.
The inspectors evaluated the licensees process for problem identification,
characterization, prioritization, and assessed whether problems were entered into the
CAP and resolved. For repetitive deficiencies and/or significant individual deficiencies in
problem identification and resolution, the inspectors verified that the licensees self-
assessment activities were capable of identifying and addressing these deficiencies.
This inspection represents one sample as defined in Inspection Procedure 71121.01-5.
b.
Findings
No findings of significance were identified.
.4
High Risk-Significant, High Dose Rate High Radiation Area (HRA) and Very High
Radiation Area (VHRA) Controls
a.
Inspection Scope
The inspectors held discussions with the Radiation Protection (RP) Manager concerning
high dose rate/HRA and VHRA controls and procedures, including procedural changes
that had occurred since the last inspection, in order to assess whether any procedure
modifications did not substantially reduce the effectiveness and level of worker
protection.
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Enclosure
This inspection constitutes one sample as defined in Inspection Procedure 71121.01-5.
The inspectors discussed with RP supervisors the controls that were in place for special
areas that had the potential to become VHRAs during certain plant operations, to
determine if these plant operations required communication beforehand with the RP
group, so as to allow corresponding timely actions to properly post and control the
radiation hazards.
This inspection constitutes one sample as defined in Inspection Procedure 71121.01-5.
The inspectors conducted plant walkdowns to assess the posting and locking of
entrances to high dose rate HRAs, and VHRAs.
This inspection constitutes one sample as defined in Inspection Procedure 71121.01-5.
b.
Findings
No findings of significance were identified.
.5
Radiation Worker Performance
a.
Inspection Scope
The inspectors reviewed radiological problem reports for which the cause of the event
was due to radiation worker errors to determine if there was an observable pattern
traceable to a similar cause, and to determine if this perspective matched the corrective
action approach taken by the licensee to resolve the reported problems. Problems or
issues with planned and taken corrective actions were discussed with the RP Manager
This inspection represents one sample as defined in Inspection Procedure 71121.01-5.
b.
Findings
No findings of significance were identified.
.6
Radiation Protection Technician Proficiency
a.
Inspection Scope
The inspectors reviewed radiological problem reports for which the cause of the event
was RP technician error to determine if there was an observable pattern traceable to a
similar cause, and to determine if this perspective matched the corrective action
approach taken by the licensee to resolve the reported problems.
This inspection represents one sample as defined in Inspection Procedure 71121.01-5.
b.
Findings
No findings of significance were identified.
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Enclosure
4.
OTHER ACTIVITIES
Cornerstone: Mitigating Systems
4OA2 Identification and Resolution of Problems (71152)
.1
Selected Issue Follow-up Inspection: Maintenance of the USAR
a.
Inspection Scope
The inspectors reviewed a sample of the licensees actions with respect to updating the
USAR in accordance with 10 CFR 50.71(e). The inspectors specifically reviewed the
licensees actions which had been completed at the time of this inspection associated
with the following corrective action documents:
- CAP039449; USAR Noted Updated to Reflect Method of Evaluation in Generic Letter
(GL) 96-06 Response; and
- CR015880; USAR May Not Have Been Updated as Required for License
Amendment 184.
The above constitutes completion of one in-depth problem identification and resolution
sample.
b. Findings
Introduction: The inspectors identified one unresolved item (URI) with respect to the
licensees updating of the USAR. Specifically, the inspectors identified that the USAR
had not been updated to reflect programmatic controls implemented to maintain the
containment sump safety function.
Description: Although specific deficiencies identified in CAP038857 for the planned
USAR update for the containment sump modification were addressed in the licensees
April 19, 2007, USAR update, the licensee had not included discussion of the
programmatic controls implemented to ensure material inside containment was
controlled. Such programmatic controls were implemented as part of the containment
sump modification (DCR 3605) and supported the analyses for the modification. The
inspectors noted that the containment sump modification was performed in response to
NRC GL 2004-02, Potential Impact of Debris Blockage on Emergency Sump
Recirculation at Pressurized Water Reactors (PWRs). The GL requested licensees to
perform an evaluation of the emergency core cooling system (ECCS) and containment
spray system recirculation functions and required licensees to provide a written
response. The inspectors noted that the programmatic controls discussed in the
licensee responses could be considered part of an analysis of a new safety issue
performed at NRC request as discussed in 10 CFR 50.71(e). The programmatic
controls implemented included control of coatings, insulation, and other materials inside
containment. In addition, the licensee had committed to perform periodic sampling of
latent debris within containment to verify that analysis assumptions were being
maintained. As these programmatic controls contributed towards maintaining the
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Enclosure
containment sump recirculation safety function, the inspectors considered these controls
germane to the containment sump analyses. This issue will be tracked as a URI
pending additional NRC review of the issue. The licensee entered this issue into their
corrective action program as CR093615, GSI-191 NRC Inspection Potential Concern
Re: USAR Update. (URI 05000305/2008002-03)
4OA3 Follow-up of Events and Notices of Enforcement Discretion (71153)
.1
(Closed) LER 05000305/2005-003-00, RHR Pumps Declared Inoperable Due to
Flooding Vulnerability
On May 5, 2006 while in intermediate shutdown, the licensee declared both trains of the
RHR system inoperable due to an internal flooding vulnerability caused by the possibility
of non-seismically qualified pipe breaks during a seismic event. The licensee indicated
that the RHR pumps were not protected from non-seismically qualified pipe breaks in the
auxiliary building. The specific design criteria in the Kewaunee USAR states that
"Class I items are protected against damage from rupture of a pipe or tank resulting in
serious flooding or excessive steam release to the extent that the Class I function is
impaired." The two RHR trains are not separated in a manner that would prevent
simultaneous damage to both trains from a failure of a non-seismically qualified pipe.
Since the plant is licensed as a hot shutdown plant, and is therefore not required to
achieve cold shutdown (which would require use of the RHR system) immediately
following a seismic event, the licensee originally interpreted that the USAR design
criteria did not apply to the RHR system.
The inspectors did not agree with this licensee interpretation and as a result Region III
submitted Task Interface Agreement (TIA) 2005-10, which requested assistance from
the Office of Nuclear Reactor Regulation to resolve this issue. The TIA response
concluded that "the design basis of the RHR system must include a provision that the
trains be separated in a manner that prevents simultaneous damage to both trains from
a failure of a non-seismic pipe." Upon receipt of the results of this TIA by licensee
station management, both RHR pumps were declared inoperable. Permanent flood
barriers were immediately installed by the licensee to protect both RHR pumps in such a
manner as to remove the internal flooding vulnerability.
Based on the complexity of this issue, the inspectors determined that the licensee would
not have reasonably identified this deviation from the USAR design criteria earlier. The
inspectors also determined that this licensee conduct was not linked to present
performance and that upon notification via the response to the TIA that such a deviation
existed, licensee corrective action was appropriate and timely. The inspectors therefore
concluded that no performance deficiency existed on this issue. This LER is closed.
This inspection constitutes one sample as defined in Inspection Procedure 71153-05.
25
Enclosure
4OA5 Other Activities
Pressurized Water Reactor Containment Sump Blockage (Temporary Instruction (TI)
2515/166)
.1
Closed NRC TI 2515/166, Pressurized Water Reactor Containment Sump Blockage
a.
Inspection Scope
The inspectors reviewed the licensees implementation of commitments
documented in their September 1, 2005 (ADAMS Accession Number ML052500378)
and February 29, 2008, (ADAMS Accession Number ML080650314) responses to
Generic Letter (GL) 2004-02. The GL addresses Generic Safety Issue (GSI) 191,
Assessment Of Debris Accumulation On PWR Sump Performance. The inspectors
reviewed licensee procedures, engineering design changes, and associated analyses.
The inspection was conducted in accordance with TI 2515-166, Pressurized Water
Reactor Containment Sump Blockage.
b.
Inspection Documentation
The questions posed by TI 2515/166 and associated status are outlined below:
(1.)
Question: Did the licensee implement the plant modifications and procedure
changes committed to in their GL 2004-02 responses? List the commitments
and the actions taken to meet each commitment. List when each action to meet
each commitment was completed. State whether additional inspections are
required to ensure all commitments have been met by the plant.
Commitment: Perform modifications to containment sump.
Commitment: Perform walkdowns of containment and evaluate debris
source term.
Commitment: Perform evaluation of strainer performance.
Commitment: Perform evaluation of chemical effects.
Commitment: Perform evaluation of downstream effects.
Commitment: Determine minimum available net positive suction head
margin for the RHR pumps at switchover to sump recirculation.
Commitment: Establish programmatic controls to ensure that potential
sources of debris introduced into containment are assessed for adverse
affects.
Commitment: Reduce post-accident debris source term.
(2.)
Question: Has the licensee updated its licensing bases to reflect the corrective
actions taken in response to GL 2004-02? Licensing bases may not be updated
until the licensee fully addresses GL 2004-02 (by December 31, 2007, unless an
extension has been granted).
26
Enclosure
(3.)
Question: If the licensee or plant has obtained an extension past the completion
date of this TI, document what actions have been completed, what actions are
outstanding, and close the TI for the plant that has the extension. Items not
finished by the TI completion date can be inspected in the future using the
generic refueling outage inspection procedure.
The strainer performance analysis was in the process of being updated
to integrate results of the June 2007 flume tests. By letter dated
November 15, 2007, (ADAMS Accession Number ML073190553), the
licensee had requested an extension for updating this analysis. As
discussed in a letter dated February 29, 2008, the licensee had scheduled
this analysis to be updated by April 30, 2008.
The licensees downstream effects calculations were in the process of being
updated to reflect changes to industry evaluation guidance (Westinghouse
Pressurized Water Reactors Owners Group WCAP-16406-P, Evaluation of
Long Term cooling Considering Particulate, Fibrous and Chemical Debris in
Recirculation Fluid, Revision 1). By letter dated November 15, 2007, the
licensee requested an extension for updating these analyses. As discussed
in a letter dated February 29, 2008, the licensee had scheduled these
analyses to be updated by May 31, 2008.
The post-LOCA containment flood level analysis was being updated to
reflect the guidance outlined in NRC letters dated August 15, 2007,
(ADAMS Accession Number ML071060091) and November 21, 2007,
(ADAMS Accession Numbers ML073110269 and ML0730278) to the
Nuclear Energy Institute. The licensee had performed a preliminary
analysis to support operability. As discussed in a letter dated
February 29, 2008, the licensee had scheduled to update the analysis by
May 31, 2008. The February 29, letter also provided a discussion of the
preliminary analysis used to support operability. The inspectors considered
the preliminary analysis sufficient to support operability and no further
inspection is required.
.2
Quarterly Resident Inspector Observations of Security Personnel and Activities
a.
Inspection Scope
During the inspection period, the inspectors conducted the following observations of
security force personnel and activities to ensure that the activities were consistent with
licensee security procedures and regulatory requirements relating to nuclear plant
security. These observations took place during both normal and off-normal plant
working hours.
Multiple tours of operations within the Central Security Alarm Stations;
Tours of selected security officer response posts;
Direct observation of personnel entry screening operations within the plant's Main
Access Facility;
Barrier/gate control activities; and
Security force vehicle inspections.
27
Enclosure
These quarterly resident inspector observations of security force personnel and activities
did not constitute any additional inspection samples. Rather, they were considered an
integral part of the inspectors' normal plant status review and inspection activities.
b.
Findings
No findings of significance were identified.
4OA6 Management Meetings
.1
Exit Meeting Summary
On April 9, 2008, the inspector presented the inspection results to Mr. S. Scace, and
other members of the licensee staff. The licensee acknowledged the issues presented.
The inspector asked the licensee whether any materials examined during the inspection
should be considered proprietary. No proprietary information was identified.
.2
Interim Exit Meetings
Interim exits were conducted for:
Occupational radiation safety program for Access to Radiologically Significant
Areas with Mr. Steve Scace on February 15, 2008.
Identification and Resolution of Problems Selected Issue Follow-Up inspection
and Pressurized Water Reactor Containment Sump Blockage (Temporary
Instruction 2515/166) inspection with Mr. S. Scace on March 28, 2008.
ATTACHMENT: SUPPLEMENTAL INFORMATION
1
Attachment
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee:
S. Scace, Site Vice President
M. Crist, Plant Manager
R. Adams, Health Physicist
L. Armstrong, Site Engineering Director
M. Bernsdorf, Chemistry
T. Breene, Nuclear Licensing Manager
W. Henry, Maintenance Manager
B. Lembeck, Radiation Protection Supervisor
C. Olsen, Health Physics Supervisor
J. Ruttar, Operations Manager
D. Shannon, Health Physics Operations Supervisor
R. Steinhardt, Site Maintenance Rule Coordinator
C. Tiernan, Corporate Maintenance Rule Coordinator
S. Wood, Emergency Preparedness Manager
Nuclear Regulatory Commission
M. Kunowski, Chief, Division of Reactor Projects, Branch 5
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened 05000305/2008002-01
Scaffolding in Close Proximity to Multiple Safety-Related
Systems Affects Operability (Section 1R04)05000305/2008002-02
Failure to Follow the Provisions of Corrective Action
Procedure PI-KW-200 Following Surveillance Testing of
containment Isolation Valve LOCA-31 (Section 1R19)05000305/2008002-03
Containment Sump Programmatic Controls Not In USAR
(Section 4OA2)
Closed 05000305/2005003-00
LER
Residual Heat Removal Pumps Declared Inoperable Due to
Flooding Vulnerability (Section 4OA3)05000305/2008002-01
Scaffolding in Close Proximity to Multiple Safety-Related
Systems Affects Operability (Section 1R04)05000305/2008002-02
Failure to Follow the Provisions of Corrective Action
Procedure PI-KW-200 Following Surveillance Testing of
containment Isolation Valve LOCA-31 (Section 1R19)
2
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 Protection
Issued Reports:
- Kewaunee USAR; Section 2.6; Hydrology; Drawing E-350; Plan - Plant Site Underground
Conduit and Cable Routes; Revision AS
- Kewaunee USAR; Section 2.6; Hydrology; Drawing E-351; Underground Conduit - Trans.
Area; Revision H
- Kewaunee USAR; Section 2.6; Hydrology; Drawing E-352; Sections and Details
Underground Conduit - Trans. Area; Revision F
- Kewaunee USAR; Section 2.6; Hydrology; Drawing 237127A-E3137; Plan and
Sections - Underground Conduit Run from Screenhouse to Diesel Room; Revision D
Procedures:
- GNP-12.06.01; Hot and Cold Weather Operations; Revision 6
- OP-KW-AOP-GEN-004; Response to Natural Events; Revision 0
- 50.59 Applicability Review of OP-KW-AOP-GEN-004; Response to Natural Events;
Revision 0
- PMP-08-19; FP - Inspection of Plant and Fire Doors; Revision 17
Work Orders:
- CR 091233; While Performing PMP-08-19 on Door 75 Inspection Revealed Torn and
Ragged Rubber Weather Stripping on North Side Near Bottom Half of Door Frame
- CR 091234; While Performing PMP-08-19 on Door 76 Inspection Revealed Torn and
Ragged Rubber Weather Stripping on Top Frame of Door
- CR 091235; While Performing PMP-08-19 on Door 437 Inspection Revealed Weather
Stripping Between the Double Doors Coming Loose - Metal Strip that Holds Weather
Stripping On is Missing Screws and is Loose
1R04 Equipment Alignment
Issue Reports:
- Current Service Water WO Tracking Search
- Drawing M-202-1; Flow Diagram Service Water System; Revision CL
- Drawing M-202-2; Flow Diagram Service Water system; Revision CS
- Drawing M-205; Flow Diagram Feedwater System; Revision BA
- Drawing XK-100-28; Flow Diagram Safety Injection System; Revision AM
- Drawing XK-100-29; Flow Diagram Safety Injection System; Revision AB
- Service Water System Health Rating Sheet
- Service Water System Health Report from 4th Quarter 2007
Procedures:
- GMP-127; Requirements and Guidelines for Scaffold Construction and Inspection;
Revisions 17 and 18
3
Attachment
- N-EHV-39; 4160V AC Supply and Distribution System Operation; Revision 24
- N-FW-05B-CL; Auxiliary Feedwater System Prestartup Checklist; Revison 40
- N-SI-33-CL; Safety Injection System Prestartup Checklist; Revision AK
- N-SW-02-CL; Service Water System Prestartup Checklist; Revision 52
- SP-42-322B; BUS 1-6 Auto Inhibit Relay Test Electrical Maintenance; Revision 10
Work Orders:
- CR 018036; Inadvertently Lifted Relief Valve SA 2050 A-1-R
- CR 027377; NRC Question Related to Turbine-Driven Auxiliary Feedwater Steam Lines in
Turbine Building
- CR 038722; Safety-Related Area Scaffold not Conforming to GNP-127 for Hot Shutdown
Mode
- CR 092303; Scaffolds Erected within 2 Inches of Safety-Related Equipment without
Engineering Evaluation/Approval
- CR 092776; Scaffolding Built within 2 Inches of Auxiliary Feedwater Trains A and B Local
Flow Indicating Piping
- CR 092791; Scaffolding Built in Contact with Air Lines to Actuators for RC-413 and RC-423
- CR 092794; Scaffolding Built Near Turbine-Driven Auxiliary Feedwater Steam Supply Piping
in Turbine Basement not Seismic
- CR 092809; Scaffolding in Auxiliary Feedwater Pump B Area Needs Further Evaluation
- CR 092901; Scaffolds Erected within 2 Inches of Safety-Related Equipment Without
Engineering Evaluation/Approval
- CR 092977; Scaffold MO1-08-095 not Constructed in Accordance with GMP-127
1R05 Fire Protection
Issued Reports:
- Active Fire Protection System Impairment Form 08-014; RTB-14 is Operable However the
Light is Obstructed Due to Scaffolding to Support DCR 3663
- Active Fire Protection System Impairment Form 08-012; The Fire Sprinkler System on the
586 Elevation of the TSC has Partial Blockage of Sprinkler Heads due to the Installation of
- Active Fire Protection System Impairment Form 08-008; Fire Suppression Sprinkler System
(heads) on the 586 Elevation of the Turbine Building West of the 1A and 1B Condensers are
being Blocked by Scaffold Decking and Asbestos Removal Tenting
- Active Fire Protection System Impairment Form 08-006; Appendix R Emergency Light
RTB-11 Located Above Door #5 on the North Wall of the Cardox Tank Room is being
Partially Obstructed by Scaffolding
- Active Fire Protection System Impairment Form 08-007; Fire Suppression Sprinkler System
(heads) on the 606 Elevation of the Turbine building Near Column Lines E and Feedwater
Heaters 14A and 14B are being Blocked by Scaffold Decking and Asbestos Removal
Tenting
- Active Fire Protection System Impairment Form;08-003; Fire Suppression Sprinkler Heads
System in the 1B Auxiliary Feedwater Pump Room are Partially (minimally) Blocked by
Scaffold Decking
- Active Fire Protection System Impairment Form 07-081; Appendix R Lighting is
Non-Functional in Zones AX-23A, AX-24, TU-92 and TU-95C
- Active Fire Protection System Impairment Form 07-091; Smoke Detector 1101-1, Located in
the Screen House Tunnel, is in Trouble Alarm
- Active Fire Protection System Impairment Form 07-095; Appendix R Light RAO2 Determined
to be Out-of-Service Due to Low Water Level and Fast Charge Indication
4
Attachment
- Active Fire Protection System Impairment Form 07-096; Non-Appendix R Light NRAMF1
Found to be Out-of-Service During Performance of PMP-41-06B
- Active Fire Protection System Impairment Form 07-100; Scaffold is Blocking Appendix R
Light EC-RAM-24
- Active Fire Protection System Impairment Form 07-104; Appendix R Emergency Light
RTB-11 Found to be Performing Incorrectly During PMP-41-06B
- Active Fire Protection System Impairment Form 07-118; Appendix R Emergency Light
RAM-10 Located Above Door #77 Near the Steam Generator Blow Down Tank is Being
Obstructed by Scaffolding and Asbestos Removal Tenting
- Active Fire Protection System Impairment Form 07-119; Appendix R Emergency Light
RAM-7, Located on the North Wall of the CST-RMST Room, is Non-functional
- Active Fire Protection System Impairment Form 06-141; Cable Spreading Room Sprinkler
System - Lack of Suppression Coverage on Certain Appendix R Cable Trays
Work Orders:
- CA 018152; 50.59 May Be Needed for Scaffold Construction in North Penetration Room
- CR 020848; 50.59 May Be Needed for Scaffold Construction in North Penetration Room
- 50.59 Applicability Review for CR 020848; 50.59 May Be Needed for Scaffold Construction
in North Penetration Room
1R11 Licensed Operator Requalification Program
Issued Reports:
- LRC-08-DY101; Cycle 08-01 Dynamic Evaluation; Revision B
1R12 Maintenance Effectiveness
Issued Reports:
- Kewaunee Power Station NRC CAP Request Data; February 11, 2008
- Kewaunee Power Station NRC CR Request Data; February 11, 2008
- Kewaunee Power Station USAR; Table 5.2-3; Reactor Containment Vessel Penetrations;
Revision 20
- Kewaunee Power Station WO Overview Report; March 12, 2008
- Kewaunee Power Station WO Overview Report - System 21; February 11, 2008
- Maintenance Rule Scoping Questions; System 21 Spent Fuel Pool Cooling System;
February 11, 2008
- Maintenance Rule System Basis; Spent Fuel Pool Cooling System; Revision 2
- Maintenance Rule System Basis; Containment Isolation; Revision 4
- Containment Isolation Report Data - September, 2006 through February, 2008
- Spent Fuel Pool Cooling Report Data - July, 2006 through December, 2007
Work Orders:
- CA 068798; Document the Spent Fuel Pool Heatup Rate
- CR 091596; NRC Resident Questions with Respect to Spent Fuel Pump Pool Maintenance
Plan
- MRE 001065; Spent Fuel Pump A Tripped Off
- MRE 001127; LOCA-3A Failed the SP-55-167-4B Post LOCA Valves Timing Test and
Needs to be Repaired
- MRE 002949; Perform a Maintenance Rule Evaluation on WR 06-3684; PEN 15 HLS
RC-422 Failed LLRT
5
Attachment
1R13 Maintenance Risk Assessments and Emergent Work Control
Issued Reports:
- Emergent Work Risk Evaluation Data; January 15, 2008
- Emergent Work Risk Evaluation Data; January 16, 2008
- Emergent Work Risk Evaluation Data; January 20, 2008
- Emergent Work Risk Evaluation Data; January 21, 2008
- Emergent Work Risk Evaluation Data; January 22, 2008
- Emergent Work Risk Evaluation Data; February 25, 2008
- Emergent Work Risk Evaluation Data; February 26, 2008
- Kewaunee Power Station Maintenance Rule 10 CFR 50.65(a)(4) Risk Projection for Week
Starting January 14, 2008
- Kewaunee Power Station Maintenance Rule 10 CFR 50.65(a)(4) Risk Projection for Week
Starting February 25, 2008
Work Orders:
- CA069790; Operations to Generate and perform an Operability Stand Down
- CR 090753; NRC Residents have Concerns with Assessing Risk of Scaffolding and Heavy
Loads
- CR 091924; Diesel Generator A Load Spiked above Limit During Loading per
OP-KW-OSPDGE-003A
- CR 092231; NRC Raises Concerns about Operability Basis of CR 091924
1R15 Operability Evaluations
Issued Reports:
- Calculation/Evaluation C11157; Auxiliary Building Basement Post Accident Area Heat Gain;
Revision Original
- Kewaunee Nuclear Power Plant Auxiliary Building Fan Level Floor EQ Equipment Data;
Revision 0
- Kewaunee Nuclear Power Plant CAP List Data; CAPs Generated on Zone SV Boundary
Issues Since March 1, 2007
- Kewaunee Nuclear Power Plant Emergency Diesel Generator 1B Largest Excursion Data;
February 8, 2007
- Kewaunee Nuclear Power Plant Emergency Diesel Generator 1B Largest Excursion Data;
February 10, 2007
- Kewaunee Nuclear Power Plant Emergency Diesel Generator 1B Largest Excursion Data;
March 6, 2008
- Kewaunee Nuclear Power Plant Engineering Log; Thursday, September 13, 2007
- Kewaunee Nuclear Power Plant Diesel Generator 1A KW Single Point Trend Analog Data;
February 28, 2008
- Kewaunee Nuclear Power Plant Design Change Request 3260; Remove Auxiliary
Feedwater Pump Suction Strainers; November 28, 2001
- Kewaunee Nuclear Power Plant Licensee Event Report AO 75-20; During Unit Startup
Operations Reduced Auxiliary Feedwater Flow was Noted with Pumps 1A and 1B in
Operation; November 15, 1975
- Kewaunee Nuclear Power Plant; Major Changes with Revision 14 of GNP-08.21.01 Data
- Kewaunee Nuclear Power Plant Root Cause Evaluation RCE 01-003; Auxiliary Feedwater
Pump Suction Strainer Configuration Not as Expected; January 23, 2001
- Kewaunee Nuclear Power Plant Safety Evaluation; Original Plant Licensing Documentation;
AFW-CST Interface; July 24, 1972
6
Attachment
- Kewaunee Nuclear Power Plant Standing Order 07-24; Requirement to Maintain Three
Auxiliary Building Basement Fan Coil Units Functional; Revision 1
- Wisconsin Public Service Corporation Correspondence; Abnormal Occurrence
Report AO 75-20; November 14, 1975
- Drawing M-704; Zone SV Exhaust System;
Procedures:
- E-0; Reactor Trip or Safety Injection; Revision 34
- 50.59 Applicability Review of E-0; Reactor Trip or Safety Injection; Revision 34
- FPP-08-09; Barrier Control; Revision 12
- GMP-208; The Opening and Sealing of Penetration Seals; Revision K
- GMP-243; Inspection and Testing of Overload Relay Heaters Electrical Maintenance
- OP-KW-ORT-DGM-001A; Emergency Diesel Generator 1A Operation Log; Revision 2
- OP-KW-OSP-DGE-003A; Operations Surveillance Procedure; Revision 1
- PMP-08-19; FP-Inspection of Fire Doors; Revision 14
- PMP-08-33; FP-Penetration Fire Barrier Inspection; Revision L
- PMP-14-02; ASV-Damper Maintenance; Revision 14
- PMP-17-02; ACA-QA-1 and QA-2 Fan Coil Units, Inspection and Cleaning; Revision 25
- SP-14-026A; Auxiliary Building Special Ventilation Train A Operability Test; Revision I
- SP-14-026B; Auxiliary Building Special Ventilation Train B Operability Test; Revision I
- SP-14-026C; Auxiliary Building Special Ventilation Train A (ASV) Monthly Test; Revision C
- SP-14-026D; Auxiliary Building Special Ventilation Train B (ASV) Monthly Test; Revision B
- SP-14-117A; Auxiliary Building Special Vent System Test Train A; Revision A
- SP-14-117B; Auxiliary Building Special Vent System Test Train B; Revision A
- SP-14-156; SV Access Door Interlock Operability Test; Revision J
- SP-24-107B; SBV Train B Operability Test; Revision M
- SP-24-107D; SBV Train B Monthly Test; Revision A
Work Orders:
- ACE 003431; SBV Train B Inoperable
- CA 010838; Licensing to Validate/Document the Licensing Basis for the Condensate Supply
- CA 015942; Auxiliary Building Basement Fan Coil Unit Operating Procedures are Non-
Conservative
- CA 016849; Auxiliary Building Basement Heat Load Calculations are Non-Conservative
- CA 029686; Diesel Generator B Exceeds 2800KW During SP-42-312B
- CA 029687; Diesel Generator B Exceeds 2800KW During SP-42-312B
- CA 031186; Diesel Generator B Exceeds 2800KW During SP-42-312B
- CA 031240; Zone SV USAR Allowed Leakage Area May Be Non-Conservative
- CA 031241; Zone SV USAR Allowed Leakage Area May Be Non-Conservative
- CA 031969; Zone SV USAR Allowed Leakage Area May Be Non-Conservative
- CA 032005; Material Stored Leaning on and next to Ductwork that is Part of Zone SV and
- CA 032196; Vendor Inspection of Injector Control Shaft Bearings from Emergency Diesel
Generator 1B
- CA 032197; Diesel Generator B Exceeds 2800KW During SP-42-312B
- CA 032237; Evaluate Methods to Control Elevator Doors as Open Barriers
- CA 032238; Revise USAR Regarding Elevator Doors
- CA 032242; SBV Train B Inoperable
- CA 032372; Disposition of Calculations C100235 and C11688
- CA 068628; Benchmark Other Sites Related to Heat Exchange Inspection and Cleaning in
Lieu of Inspection
7
Attachment
- CA 068629; Engineering Program - Inspection and Material to Capture Documentation
within a Procedure
- CA 069790; NRC Raises Concerns About Operability Basis of CR 091924
- CE 020244; NRC Resident Inspector Questioned if Elevator Doors are Zone SV Boundaries
- CAP 041567; Diesel Generator B Exceeds 2800KW During SP-42-312B
- Apparent Cause Evaluation 3374 for CAP 041567; Diesel Generator B Exceeds 2800KW
During SP-42-312B
- CAP 043792; NRC Resident Inspector Questioned if Elevator Doors are Zone SV
Boundaries
- CAP 043818; Zone SV USAR Allowed Leakage Area May Be Non-Conservative
- CAP 044013; BAST Room Floor Drain Open to Non-SV/Non-Steam Exclusion Area
- CAP 044432; SBV Train B Inoperable
- Apparent Cause Evaluation of CAP 044432; SBV Train B Inoperable
- CAP 044796; Material Stored Leaning on and next to Ductwork that is Part of Zone SV and
- CE 020246; Zone SV USAR Allowed Leakage Area May Be Non-Conservative
- CR 012915; Auxiliary Building Mezzanine Fan Coil Unit B Air Flow is Lower than Expected
- CR 013788; NRC Resident Concern on Non-Safety to Safety Interface condensate to
Auxiliary Feedwater System
- CR 090907; Documentation of Kewaunee Power Station Justification for Heat Exchange
Inspection/Cleaning in Lieu of Testing
- CR 019147; Auxiliary Building Basement Heat Load Calculations are Non-Conservative
- CR 019674; C11147 Auxiliary Building Fan Floor Heat Gain Calculation is Non-Conservative
- CR 019676; Auxiliary building Fan Floor Heat Gain Calculation has Inadequate Technical
Basis
- RAS 39, Revision 1 of CR 019676; Auxiliary Building Fan Floor Heat Gain Calculation has
Inadequate Technical Basis
- CR 020597; Incorrect Assumption Made in Fan Floor Heat Up Evaluation
- CR 029317; BT-32B Exceeded the Action Limits for Closing and Opening During Retest
- CR 029326; Problems Discovered with Replacement Asco Solenoid Valve
- CR 091907; Emergency Diesel Generator Governor Oil Level Information Transmittal
- CR 091924; Diesel Generator A Load Spiked Above Limit During Loading Per
OP-KW-OSP-DGE-003A
- CR 092231; NRC Raises Concerns About Operability Basis of CR 091924
- KW 07-001462; Diesel Generator B Load Swings During Run on 07
- KW 100307473; Open, Inspect Available Tubes with Boroscope and Backflush 1D Auxiliary
Building Basement Fan coil Unit
- MRE003047; Diesel Generator B Exceeds 2800KW During SP-42-312B
- MRE 003088; SBV Train B Inoperable
- WO 07-006318-000; SBV Train B Failed to Start During SP-24-107D
1R18 Plant Modifications
Issued Reports:
- Edward Alsteen/NonGasLDC/VANCP OWER Correspondence; Transformer B Bay Deluge
Piping Support Removal; October 6, 2007
Procedures:
- FP-E-MOD-03; Temporary Modifications; Revision 0
- MA-AA-101; Rigging Lift Plan; Revision 1
- VPAP-1403; Temporary Modifications; Revision 11
8
Attachment
- Modification 3631-1; Generator Step-Up Transformer Replacement; Revision 0
Work Orders:
- DCR 3631-1; Generator Step-Up (GSU) Transformer Replacement
- 50.59 Applicability Review of DCR 3631-1; Generator Step-Up (GSU) Transformer
Replacement
- 07-001436-000; Remove the Pre-cast Concrete Half-Walls in Front of the Main Transformer
Bays and the Main Transformer Spare Bay
1R19 Post-Maintenance Testing
Issued Reports:
- Machine 1B Water Pump; Last Measurement Report Data; February 8, 2008
- Nuclear Management Company Correspondence to Nuclear Regulatory Commission;
Application for Technical Specification Improvement to Eliminate Requirements for Hydrogen
Recombiners and Hydrogen/Oxygen Monitors; January 30, 2004
- Nuclear Regulatory Commission Correspondence to Nuclear Management Company;
May 13, 2004; Issuance of Amendment Regarding Relocation of Requirements for Hydrogen
Monitor
- Nuclear Regulatory Commission Federal Register, Volume 67, No. 149; RIN 3150-AG76;
Combustible Gas Control in Containment; August 2, 2002
- Nuclear Regulatory Commission Federal Register, Volume 68, No. 186; 67 FR 50374;
Relax the Hydrogen and Oxygen Monitor Requirements; September 25, 2003
Procedures:
- GMP-131; Operational Use for SKF Microlog Analyzers; Revision G
- GNP-01.09.01; Service Water and Fire Protection System Inspection program and
Coordination; Revision C
- GNP-03.30.06; Plant Status and Configuration Control; Revision 8
- GNP-04.04.01; 50.59 Applicability Review and Pre-Screening; Revision K
- MA-KW-ICP-MS-001A; Steam Generator A Power Operated Relief Valve and Control Loop
Calibration and SD-3A Trip Valve Rebuild; Revision 1
- MA KW-ICP-SW-071B2; Service Water Pump 1B2 Lube Water Pressure Regulator
Maintenance; Revision 0
- 59.59 Applicability Review for MA KW-ICP-SW-071B2; Service Water Pump 1B2 Lube
Water Pressure Regulator Maintenance; Revision 0
- 59.59 Applicability Review for MA KW-ICP-SW-071B2; Service Water Pump 1B2 Lube
Water Pressure Regulator Maintenance; Revision 1
- OP-AA-102; Operability Determination; Revision 0
- OP-AA-102-1001; Development of Technical Basis to Support Operability; Revision 0
- OP-KW-ORT-SW-002B; Service Water Pump Train B Backup Bearing Lube Water Supply
Check; Revision 0
- OP-KW-OSP-DGE-002A; Diesel Generator A Quarterly Availability Test; Revision 1
- PI-AA-300; Cause Evaluation; Revision 1
- PI-KW-200; Corrective Action; Revision 3
- PMP-17-02; ACA-QA-1 & QA-2 Fan Coil Unites - Inspection and Cleaning; Revision 25
- SP-55-167-4B; Post LOCA Valves Timing Test (IST) from Local Panel - Train B; Revision B
Work Orders:
- ACE 000768; SD-3A Opened Fully when MS-1A was Closed
- Apparent Cause Evaluation for ACE 000768; SD-3A Opened Fully when MS-1A was Closed
9
Attachment
- ACE 013652; Timing Test for LOCA-3A Exceeded Action Values
- CA 022013; LOCA-3A Opening Time Near Action Value
- CA 068628; Documentation of Kewaunee Power Station Justification for Heat Exchange
Inspection/Cleaning in Lieu of testing
- CA 068629; Documentation of Kewaunee Power Station Justification for Heat Exchange
Inspection/Cleaning in Lieu of testing
- CR 019147; RAS 37 Auxiliary Basement Heat Load Evaluation
- CR 025595; LOCA-3A Opening Time Near Action Value
- CR 028605; LOCA-3A Failed the SP-55-167-4B Post LOCA Valves Timing Test (IST) and
Needs to be Repaired
- Apparent Cause Evaluation 918 of CR 028605
- Apparent Cause Evaluation 919 of CR 028605
- CR 028647; Containment Hydrogen Monitor A Nonfunctional
- Apparent Cause Evaluation ACE00916 of CR 028647
- CR 090000; LOCA-3A Closed Limit Switch Unable to be Adjusted to GIP-020A Specs
- CR 090002; LOCA-3A Closed Limit Switch Unable to be Adjusted to GIP-020A Specs
- CR 090006; LOCA-3A Remains Inoperable Following Actuator Overhaul - Failed Timing
Test
- CR 090616; Out of Specification as Found Reading while Performing
MA-KW-ICP-SW-071A2
- CR 090907; Documentation of Kewaunee Power Station Justification for Heat Exchange
Inspection/Cleaning in Lieu of Testing
- CR 093059; Conn Code on Spare Foxboro Box Incorrect for Internal Wiring
- CR 093066; Power Cord to PC-468A Making Poor Connection to the Controller
- KW 07-011591; Rebuild or Replace Service Water 1B2 Regulator
- KW-100307473; Open, Inspect Available Tubes with Boroscope and Backflush 1D Auxiliary
Building Basement Fan Coil Unit
- KW-100309607; LOCA-3A Opening Time Near Action Value
- KW 100341690; SD-3A Controller Output
- WO 06-11479-000; Plant Equipment Water Pump B Motor is Chirping
1R20 Outage Activities
Procedures:
- N-CRD-49 R-27; Control Rod Drive
- N-HB-11 R-25; Heater and Moisture Separator-Drain Bleed Steam System
- N-TB-54 R-80; Turbine and Generator Operation
- OP-KW-GOP-206 R-1; Shutdown from Full Power to 35% Power
1R22 Surveillance Testing
Issued Reports:
- Calculation/Evaluation C11157; Auxiliary Building Basement Post Accident Area Heat Gain;
Revision Original
- Diesel Generator B Performance Indicator Data; January 10, 2008
- Emergency Diesel Generator 1B Operation Log; January 10, 2008
- Foreign Material Exclusion Evaluation of SP-55-155A
- Kewaunee Nuclear Power Plant Auxiliary Building Fan Level Floor EQ Equipment Data;
Revision 0
- Kewaunee Nuclear Power Plant CAP List Data; CAPs Generated on Zone SV Boundary
Issues Since March 1, 2007
10
Attachment
- Kewaunee Nuclear Power Plant Engineering Log; Thursday, September 13, 2007
- Train B Automatic Load Sequencer Test; January 10, 2008
- Kewaunee Nuclear Power Plant Standing Order 07-24; Requirement to Maintain Three
Auxiliary Building Basement Fan Coil Units Functional; Revision 1
- Drawing M-704; Zone SV Exhaust System;
Procedures:
- E-0; Reactor Trip or Safety Injection; Revision 34
- 50.59 Applicability Review of E-0; Reactor Trip or Safety Injection; Revision 34
- FPP-08-09; Barrier Control; Revision 12
- GMP-208; The Opening and Sealing of Penetration Seals; Revision K
- GMP-243; Inspection and Testing of Overload Relay Heaters Electrical Maintenance
- OP-KW-OSP-DGE-001A; Diesel Generator A Monthly Availability Test; Revision 2
- OP-KW-OSP-DGE-001B; Diesel Generator B Monthly Availability Test; Revision 2
- PMP-08-19; FP-Inspection of Fire Doors; Revision 14
- PMP-08-33; FP-Penetration Fire Barrier Inspection; Revision L
- PMP-14-02; ASV-Damper Maintenance; Revision 14
- PMP-17-02; ACA-QA-1 and QA-2 Fan Coil Units, Inspection and Cleaning; Revision 25
- SP-14-026A; Auxiliary Building Special Ventilation Train A Operability Test; Revision I
- SP-14-026B; Auxiliary Building Special Ventilation Train B Operability Test; Revision I
- SP-14-026C; Auxiliary Building Special Ventilation Train A (ASV) Monthly Test; Revision C
- SP-14-026D; Auxiliary Building Special Ventilation Train B (ASV) Monthly Test; Revision B
- SP-14-117A; Auxiliary Building Special Vent System Test Train A; Revision A
- SP-14-117B; Auxiliary Building Special Vent System Test Train B; Revision A
- SP-14-156; SV Access Door Interlock Operability Test; Revision J
- SP-24-107B; SBV Train B Operability Test; Revision M
- SP-24-107D; SBV Train B Monthly Test; Revision A
- SP-31-168B; Train B Component Cooling Pump and Valve Test - IST; Revision 15
- SP-45-049.21; RMS Channel R-21 Containment Stack Radiation Monitor Quarterly
Functional Test; Revision U
- SP-55-155A; Engineered Safeguards Train A Logic Channel Test; Revision 25
- SP-55-167-4A; Post LOCA Valves Timing Test (IST) from Local Panel - Train A; Revision B
- SP-55-167-4B; Post LOCA Valves Timing Test (IST) from Local Panel - Train B; Revision B
Work Orders:
- ACE 003431; SBV Train B Inoperable
- CR 012915; Auxiliary Building Mezzanine Fan Coil Unit B Air Flow is Lower than Expected
- CA 015942; Auxiliary Building Basement Fan Coil Unit Operating Procedures are Non-
Conservative
- CA 016849; Auxiliary Building Basement Heat Load Calculations are Non-Conservative
- CA 016879; Auxiliary Building Basement Heat Load Calculations are Non-Conservative
- CA 032005; Material Stored Leaning on and next to Ductwork that is Part of Zone SV and
- CA 031240; Zone SV USAR Allowed Leakage Area May Be Non-Conservative
- CA 031241; Zone SV USAR Allowed Leakage Area May Be Non-Conservative
- CA 031969; Zone SV USAR Allowed Leakage Area May Be Non-Conservative
- CA 032237; Evaluate Methods to Control Elevator Doors as Open Barriers
- CA 032238; Revise USAR Regarding Elevator Doors
- CA 032242; SBV Train B Inoperable
- CA 032372; Disposition of Calculations C100235 and C11688
11
Attachment
- CA 068628; Benchmark Other Sites Related to Heat Exchange Inspection and Cleaning in
Lieu of Inspection
- CA 068629; Engineering Program - Inspection and Material to Capture Documentation
within a Procedure
- CE 020244; NRC Resident Inspector Questioned if Elevator Doors are Zone SV Boundaries
- CE 020246; Zone SV USAR Allowed Leakage Area May Be Non-Conservative
- CAP 043792; NRC Resident Inspector Questioned if Elevator Doors are Zone SV
Boundaries
- CAP 043818; Zone SV USAR Allowed Leakage Area May Be Non-Conservative
- CAP 044013; BAST Room Floor Drain Open to Non-SV/Non-Steam Exclusion Area
- CAP 044432; SBV Train B Inoperable
- Apparent Cause Evaluation of CAP 044432; SBV Train B Inoperable
- CAP 044796; Material Stored Leaning on and next to Ductwork that is Part of Zone SV and
- CR 090907; Documentation of Kewaunee Power Station Justification for Heat Exchange
Inspection/Cleaning in Lieu of Testing
- CR 019147; Auxiliary Building Basement Heat Load Calculations are Non-Conservative
- CR 019674; C11147 Auxiliary Building Fan Floor Heat Gain Calculation is Non-Conservative
- CR 019676; Auxiliary Building Fan Floor Heat Gain Calculation has Inadequate Technical
Basis
- RAS 39, Revision 1 of CR 019676; Auxiliary Building Fan Floor Heat Gain Calculation has
Inadequate Technical Basis
- CR 020597; Incorrect Assumption Made in Fan Floor Heat Up Evaluation
- KW 07-011268; PM55-001 Monthly Test
- KW 100307473; Open, Inspect Available Tubes with Boroscope and Backflush 1D Auxiliary
Building Basement Fan Coil Unit
- MRE 003088; SBV Train B Inoperable
- WO 07-006318-000; SBV Train B Failed to Start During SP-24-107D
1EP6 Drill Evaluation
Issued Reports:
- LRC-08-DY101; Cycle 08-01 Dynamic Evaluation; Revision B
2OS1 Access Control to Radiologically Significant Areas
Issued Reports:
- Audit 07-06; Radiological Protection, Process Control Program, and Chemistry Programs;
July 26, 2007
Procedures:
- RE-24; Special Nuclear Materials Control; Revision P
- HP-01.021; Issuance and Control of Locked High Radiation Keys; Revision F
- HP-03.006; In-Vitro Bioassay Measurement; Revision F
- HP-05.022; Controls for Transfer of Radioactive Material; Revision 4
- RP-AA-202; Radiological Posting; Revison 0
- RP-KW-03-008; Evaluation of Inhalation or Ingestions; Revision 0
- RP-KW-03-009; Calculating Internal Dose from Whole Body Counter Results; Revision 0
- RP-KW-001-024; Posting and Shielding Guidance for Fuel Movement at KPS; Revision 0
- RP-KW-005-005; Radiation and Contamination Survey and Airborne Radioacitivity Sampling
Schedules; Revision 0
12
Attachment
Work Orders:
- CAP 042477; Security Force Member Entered RCA with Lunch Box
- CR 016137; Higher than Expected Dose Rate not Reported to On Shift RP Technician
- CR 0196766; Procedure not Followed for Issuance of Respirator
- CR 023925; Security Force member Received Dose of 14 Mrem in Auxiliary Building
- CR 025085; Performing a Source Check on R-23 Disables Alarms
- CR 025939; Document the Dose Delta for the Change Out of the Letdown Bag Filter
- CR 025101; Missed Shielding Walkdown
- CR 091008; Procedure HP-01.021 and RP-KW-001-004 Wording Differed from the
- CR 091086; Inventory of Locked High Radiation Area Keys not Completed for the
Emergency Annulus Keys
- CR 091010; Locked High Radiation Area Key Inventory Enhancements
4OA1 Performance Indicator Verification
Issued Reports:
- Performance Indicator Data Sets, Service Water; January, 2007 - December, 2007
- Performance Indicator Data Sets, Diesel Generators; January, 2007 - December, 2007
- Performance Indicator Data Sets, Component Cooling; January, 2007 - December, 2007
- Performance Indicator Data Sets, Safety Injection; January, 2007 - December, 2007
- Performance Indicator Data Sets, Residual Heat Removal; January, 2007 - December, 2007
4OA2 Problem Identification and Resolution
Procedures:
- NEP-05.02; Revision and Control of the Updated Safety Analysis Report; Revision 7,
Work Orders:
- CAP038857; USAR Revision for DCR 3605; dated October 27, 2006
- CAP039449; USAR Not Updated to Reflect Method of Evaluation in GL 96-06 Response;
dated November 16, 2006
- CR015880; USAR May Not Have Been Updated as Required for License Amendment 184;
dated July 13, 2007
- CR093615; GSI-191 NRC Inspection Potential Concern Re: USAR Update; dated
March 24, 2008 [NRC Identified]
4OA3 Follow-up of Events and Notices of Enforcement Discretion
Issued Reports:
- Event Notification 44027; Planned maintenance on Mishicot Substation by Wisconsin Public
Service Results in Greater than 50% siren Coverage Loss; March 4, 2008
- Control Room Shift turnover Checklist of February 19, 2008
Procedures:
- OP-KW-ARP-47065-0; Condenser Hotwell Level High/Low; Revision 0
Work Orders:
- CA 069037; Operations for CR 091246 to Track Completion of the MU-3B Alternate Plant
Configuration
13
Attachment
- CR 091245; Documenting Alternate Plant configuration that was Created Due to an Issue
with Main Condenser Hotwell Level Indicator L24011
- CR 091246; Alternate Plant Configuration for MU-3B Line Due to Level Instrument Issue
4OA5 Other Activities
Calculations:
- 51-9017897; Kewaunee RHR, SI and ICS Pump Evaluation for GSI-191 Downstream Effects
[Proprietary]; Revision 1
- 51-9014070; Kewaunee Strainer Performance Test Report; Revision 1
- 51-9020502; Chemical Precipitation Analysis for Kewaunee Power Station Using WCAP-
16530-NP; Revision 3
- 51-9054883; Kewaunee Containment Debris Trap Efficiency Test Report; Revision 1
- 2004-08820; GSI-191 Debris Generation; Revision 3
- 2004-08820; GSI-191 Debris Generation Calculation, Debris Inventory; Revision 3
Addendum A
- 2005-1400; GSI-191 Downstream Effects - Flow Clearances; Revision 0
- 2005-13160; Phase II Downstream Evaluation for Resolution of GSI-191; Revision 1
- 2006-01660; Post LOCA Containment Flood Level (DCR 3605); Revision 0
- ALION-REP-DOM-4458-02; Kewaunee High Density Fiberglass Debris Erosion Testing
Report [Proprietary]; Revision 0
- FP-E-MOD-04; Design Input Checklist (Part B - Design Considerations, Requirements, and
Standards); Revision 2
- OP-KW-GCL-102B; Plant Requirements for Exceeding 200°F; Revision 0
- OP-KW-GOP-102; Startup From Cold Shutdown to RHR; Revision 2
- PCI-5407-S01; Structural Evaluation of Containment Sump Strainers; Revision 2
- PCI-5407-S02; Evaluation of Sump Cover and Piping for the Containment Sump Strainers;
Revision 3
- TDI-6008-06; Total Head Loss (ECCS Recirculation Strainer) - Kewaunee Power Station;
Revision 7
- TDI-6008-07; Vortex, Air Ingestion & Void Fraction (ECCS Recirculation
Strainer) -- Kewaunee Power Station; Revision 3
Procedures:
- CM-AA-CRS-10; Containment Recirculation Sump GSI-191 Program; Revision 0
- CM-AA-CRS-100; GSI Program Standards, Requirements, and Guidance for the
Containment Recirculation Sump; Revision 0
- CM-AA-CRS-103; Containment Coating Condition Assessment; Revision 0
- ES-3000; Specification for Insulation - General; Revision 7
- ES-3003; Specification for Insulation - Nuclear Steam Supply System; Revision 4
- GMP-262; General Insulation Information; Revision C
- GNP-01.31.01; Plant Cleanliness and Storage; Revision 17
- GNP-08.06.02; Containment Hot Shutdown Walkdown; Revision 4
- GNP-08.22.01; Protective Coating Application for Service Level I Areas Inside the Reactor
Containment Vessel; Revision 9
- GNP-12.17.01; Cold Shutdown Containment Inspection; Revision 9
- GNP-12.17.02; Containment Inspection During Operations; Revision 9
- MA-AA-102; Foreign Material Exclusion; Revision 4
- N-CCI-56; Containment Access; Revision 21
- NAD-08.22; Protective Coatings Program; Revision 5
14
Attachment
- NEP-04.22; Containment Latent Debris Sampling Evaluation; Revision A
- NEP-04.23; Containment Latent Debris Sample Collection; Revision A
Work Orders:
- CA025943; Inappropriate Corrective Action for CAP032490; dated September 5, 2006
- CA071163; Implement Fleet Procedure Process for Safety and Non-safety Procedures the
Same; dated March 27, 2008 [NRC Identified]
- CAP038857; USAR Revision for DCR 3605; dated October 27, 2006
- CR093709; NRC Inspector Questions Procedure Classifications; dated March 25, 2008
[NRC Identified]
- LBL024275; Component Labeling; dated June 15, 2006
- Modification DCR3605; Replacement of the ECCS Sump B Strainer; Revision 3
- KW06-003290; S/G B/D Piping Insulation in Containment Basement; Revision 0
- KW06-003292; RF28 - Shroud Cooling SW Lines, Replace Insulation; Revision 0
- KW06-011598; Steam Generator Blowdown piping insulation in containment; Revision 0
15
Attachment
LIST OF ACRONYMS USED
Corrective Action Program
CFR
Code of Federal Regulations
CR
Condition Report
Division of Reactor Projects
GL
Generic Letter
Generic Safety Issue
IMC
Inspection Manual Chapter
Inservice Testing
LER
Licensee Event Report
Non-Cited Violation
NRC
U.S. Nuclear Regulatory Commission
Performance Indicator
Post-Maintenance
Pressurized Water Reactor
Radiation Protection
Significance Determination Process
Structure, System and Component
Turbine-Driven Auxiliary Feedwater
TI
Temporary Instruction
Task Interface Agreement
TS
Technical Specification
Updated Final Safety Analysis Report
Updated Safety Analysis Report
Unresolved Item
Very High Radiation Area
Work Order