ML072840428
| ML072840428 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 10/10/2007 |
| From: | Howell A NRC/RGN-IV/DRP |
| To: | Ridenoure R Omaha Public Power District |
| References | |
| EA-07-194 IR-07-011 | |
| Download: ML072840428 (26) | |
See also: IR 05000285/2007011
Text
October 10, 2007
R. T. Ridenoure
Vice President
Omaha Public Power District
Fort Calhoun Station FC-2-4 Adm.
P.O. Box 550
Fort Calhoun, NE 68023-0550
SUBJECT: FORT CALHOUN STATION NRC INSPECTION REPORT 05000285/2007011
Dear Mr. Ridenoure:
On September 18, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an
inspection at your Fort Calhoun Station. The enclosed inspection report documents the
inspection findings, which were discussed on September 18, 2007, with Mr. Tim Nellenbach,
Plant Manager, and other members of your staff. This report documents baseline inspection
activities related to Train A emergency diesel generator failures which occurred on
February 14, 2007, and February 16, 2007. Region IV management decided to document this
inspection in a separate report because the underlying performance deficiencies were complex
and the preliminary significance of associated findings appeared to be of greater than very low
safety significance. 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.
This report discusses a finding that appears to have Greater than Green (greater than very low)
safety significance. As described in Section 4OA2 of this report, contamination containing dust
and oil was found on the field flash relay auxiliary contact surfaces, which apparently caused
the February 14, 2007 emergency diesel generator failure. Our inspectors determined
that: (1) craftsmen were applying an unapproved wet lubricant to the auxiliary contact sliding
mechanisms, contrary to vendor recommendations and in the absence of procedural controls;
(2) that Fort Calhoun Station staff did not treat the February 14, 2007, emergency diesel
generator failure as a significant condition adverse to quality; and (3) actions in response to
applicable operating experience were not timely and did not prevent this condition from
occurring. These issues were assessed for safety significance based on the best available
information, including influential assumptions, using the applicable Significance Determination
Process (SDP). The NRC determined that preliminarily the above stated issues, which resulted
in two apparent violations, are cumulatively considered as one finding, which was determined to
have Greater than Green significance. The final resolution of this finding will convey the
increment in the importance to safety by assigning the corresponding color. This preliminary
finding has Greater than Green significance because the Train A emergency diesel generator
Omaha Public Power District
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was unavailable for a significant period (between 14 and 28 days, depending on failure mode
assumptions). The primary assumptions associated with the preliminary SDP are documented
in Attachment 2 to this report.
The finding is associated with two apparent violations of NRC requirements and is being
considered for escalated enforcement action in accordance with the NRC Enforcement Policy.
The current Enforcement Policy is included on the NRCs Web site at
www.nrc.gov/about-nrc/regulatory/enforcement.html
Before we make a final decision on this matter, we are providing you an opportunity (1) to
present to the NRC your perspectives on the facts and assumptions, used by the NRC to arrive
at the finding and its significance, at a Regulatory Conference or (2) submit your position on the
finding to the NRC in writing. If you request a Regulatory Conference, it should be held within
30 days of the receipt of this letter and we encourage you to submit supporting documentation
at least one week prior to the conference in an effort to make the conference more efficient and
effective. If a Regulatory Conference is held, it will be open for public observation. If you
decide to submit only a written response, such submittal should be sent to the NRC within 30
days of the receipt of this letter. In either case, please provide the following additional
information:
Your position and justification for the applicable exposure time - whether the exposure
time should be 28 days (the full duration between surveillances) or 14 days (t/2) if the
exact time the diesel became inoperable was unknown.
Your own assessment of the increase in core damage frequency associated with the
Train A emergency diesel generator being unable to perform its safety function. Your
assessment should include a discussion of the contribution of internal and external
Your views on the origins of the oil and dust contaminants that were found on the
auxiliary contact surfaces.
These issues do not represent an immediate safety concern because of the corrective actions
you have taken. These actions involved inspecting and replacing, as necessary, vulnerable
auxiliary contacts in the emergency diesel generator control circuits.
Please contact Mr. Jeffrey Clark at (817) 860-8147 within 10 business days of the date of the
receipt of this letter to notify the NRC of your intentions. If we have not heard from you within
10 days, we will continue with our significance determination and enforcement decision and you
will be advised by separate correspondence of the results of our deliberations on this matter.
Since the NRC has not made a final determination in this matter, no Notice of Violation is being
issued for the inspection finding at this time. In addition, please be advised that the number
and characterization of apparent violations described in the enclosed inspection report may
change as a result of further NRC review.
Omaha Public Power District
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In addition, this report documents one self-revealing finding of very low safety significance
(Green). This finding was determined to involve a violation of NRC requirements. However,
because of the very low safety significance and because it was entered into your corrective
action program, the NRC is treating this finding as a noncited violation (NCV), consistent with
Section VI.A.1 of the NRC Enforcement Policy. If you contest the violation or significance of the
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, D.C. 20555-0001, with copies to the Regional Administrator, U.S.
Nuclear Regulatory Commission, Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas
76011-4005; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission,
Washington, D.C. 20555-0001; and the NRC Resident Inspector at the Fort Calhoun Station
facility.
In accordance with 10 CFR Part 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 component of NRCs document
system (ADAMS). ADAMS is accessible from the NRC Web site at
"http://www.nrc.gov/reading-rm/adams.html" (the Public Electronic Reading Room).
Sincerely,
/RA/ Tony Vegel for
Arthur T. Howell III, Director
Division of Reactor Projects
Docket: 50-285
License: DPR-40
Enclosure:
NRC Inspection Report 05000285/2007011
w/Attachment: Supplemental Information
cc w/Enclosure:
Joe l. McManis, Manager - Licensing
Omaha Public Power District
Fort Calhoun Station FC-2-4 Adm.
P.O. Box 550
Fort Calhoun, NE 68023-0550
David J. Bannister
Site Director - Fort Calhoun Station
Omaha Public Power District
Fort Calhoun Station FC-1-1 Plant
P.O. Box 550
Fort Calhoun, NE 68023-0550
Omaha Public Power District
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James R. Curtiss
Winston & Strawn
1700 K Street NW
Washington, DC 20006-3817
Chairman
Washington County Board of Supervisors
P.O. Box 466
Blair, NE 68008
Julia Schmitt, Manager
Radiation Control Program
Nebraska Health & Human Services
Dept. of Regulation & Licensing
Division of Public Health Assurance
301 Centennial Mall, South
P.O. Box 95007
Lincoln, NE 68509-5007
Melanie Rasmussen
Bureau of Radiological Health
Iowa Department of Public Health
Lucas State Office Building, 5th Floor
321 East 12th Street
Des Moines, IA 50319
Omaha Public Power District
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Electronic distribution by RIV:
Regional Administrator (EEC)
DRP Director (ATH)
DRS Director (DDC)
DRS Deputy Director (RJC1)
Senior Resident Inspector (JDH1)
Resident Inspector (LMW1)
Branch Chief, DRP/E (JAC)
Senior Project Engineer, DRP/E (GDR)
Team Leader, DRP/TSS (CJP)
RITS Coordinator (MSH3)
OEMail
G. M. Vasquez, RIV
V. L. Dricks, RIV
W. M. Maier, RIV
M. a. Ashley, NRR
J. Wray, OE
Only inspection reports to the following:
D. Pelton, OEDO RIV Coordinator (DLP)
ROPreports
FCS Site Secretary (BMM)
SUNSI Review Completed: GDR ADAMS: G Yes No Initials: GDR
G Publicly Available G Non-Publicly Available G Sensitive G Non-Sensitive
R:\\_REACTORS\\_FCS\\2007\\FCS 2007-011_gdr.wpd
RIV:SPE:DRP/E
RI:DRP/E
SPE:DRP/E
C:DRS/EB1
C:DRS/OB
GDReplogle
LMWilloughby
JDHanna
WBJones
ATGody
/RA/
E-GDReplogle
T-GDReplogle
/RA/
MEMurphy for
09/28/07
09/28/07
10/02/07
09/26/07
09/26/07
C:DRS/EB2
SRA:DRS
ACES
C:DRP/E
D:DRS
LJSmith
RLBywater
GMVasquez
JAClark
DDChamberlain
/RA/
/RA/
GDReplogle for
09/27/07
10/ /07
09/30/07
10/02/07
10/02/07
D:DRP
ATHowell
AVegel for
10/10/07
OFFICIAL RECORD COPY
T=Telephone E=E-mail F=Fax
Enclosure
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U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket:
50-285
License:
Report:
Licensee:
Omaha Public Power District
Facility:
Fort Calhoun Station
Location:
Fort Calhoun Station FC-2-4 Adm.
P.O. Box 399, Highway 75 - North of Fort Calhoun
Fort Calhoun, Nebraska
Dates:
February 16, 2007 through September 18, 2007
Inspectors:
J. Hanna, Senior Resident Inspector
L. Willoughby, Resident Inspector
G. Replogle, Senior Project Engineer
Reactor Analyst
R. Bywater, Senior Reactor Analyst
Branch Chief
Jeff Clark, Chief, Project Branch E
Division of Reactor Projects
Approved By:
Arthur T. Howell, Director
Division of Rector Projects
Enclosure
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SUMMARY OF FINDINGS
IR 05000285/2007011; 02/14/2007 - 09/18/2007; Fort Calhoun Station, Resident Report;
Postmaintenance Testing, Identification and Resolution of Problems.
The report covered a 31 week period of inspection by a senior resident inspector, a resident
inspector and a senior project engineer. One Green noncited violation and two apparent
violations of significance were identified. The significance of most findings is indicated by their
color (Green, White, Yellow, or Red) using Inspection Manual Chapter 0609, Significance
Determination Process. Findings for which the significance determination process does not
apply may be Green or be assigned a severity level after NRC management review. The
NRC's program for overseeing the safe operation of commercial nuclear power reactors is
described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.
A.
NRC-Identified Findings and Self-Revealing Findings
Cornerstone: Mitigating Systems
Green. A self-revealing Green noncited violation of Technical Specification 5.8.1.a
(Procedures) was identified for an inadequate postmaintenance testing procedure.
Craftsmen had replaced the field flash relay auxiliary contacts (following a previous field
flash failure on February 14, 2007) and had misaligned the contact assembly during
installation. Postmaintenance testing was inadequate because it did not verify that the
contacts properly repositioned to the closed position following the surveillance test.
When the emergency diesel generator was started two days later, for a normal
surveillance test, the field did not flash because the contacts were stuck open.
This finding was greater than minor because the finding was associated with the
mitigating systems cornerstone objective (procedure quality attribute) to ensure the
availability, reliability, and capability of systems that respond to initiating events to
prevent undesirable consequences. The exposure time for this performance deficiency
was approximately 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />. Using the Manual Chapter 0609, Appendix A,
Determining the Significance of Reactor Inspection Findings for At-Power Situations,
Phase 1 screening worksheet, the inspectors determined that the finding was of very
low safety significance (Green) because it was not: 1) a design or qualification
deficiency; 2) a loss of system safety function; 3) an actual loss of safety function for
greater than its technical specification allowed outage time; 4) a loss of safety function
of a non-technical specification train; or 5) a seismic, flooding or severe weather related
finding. The finding had crosscutting aspects in the human performance area,
specifically the resource attribute (H.2(c)) in that a complete and accurate test
instruction was not provided to test the 2CR auxiliary relay contacts (Section 1R19).
TBD. The inspectors identified an apparent violation of 10 CFR 50, Appendix B,
Criterion XVI (Corrective Actions), with two examples, for the failure to: 1) treat the
February 14, 2007, emergency diesel generator failure as a significant condition
adverse to quality; and 2) promptly identify and correct a significant condition adverse to
quality (high resistance on field flash circuit contacts) after determining that similar
operating experience was applicable. In addition, a contributor to the inoperable
Enclosure
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emergency diesel generator included the failure to revisit the diesel generator operability
evaluation in response to the applicable operating experience. Overall, the licensee
responded to various problems in isolation and did not adopt a corrective action process
that maintained emergency diesel generator reliability and availability.
This apparent violation was greater than minor because it affected the mitigating
systems cornerstone objective (equipment performance attribute), to ensure the
availability, reliability, and capability of systems that respond to initiating events to
prevent undesirable consequences. For the preliminary significance determination, the
inspectors used a 14 day exposure time, which was half the time period between the
last successful surveillance and the February 14, 2007, failure. However, this exposure
time could increase to 28 days if the NRC determines the failure was caused by contact
binding, versus contamination. Using the NRC Inspection Manual Chapter 0609,
Appendix AProperty "Inspection Manual Chapter" (as page type) with input value "NRC Inspection Manual 0609,</br></br>Appendix A" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., Determining the Significance of Reactor Inspection Findings for At-Power
Situations, significance determination process, a Region IV senior reactor analyst
determined that the finding was potentially Greater than Green. The finding had
crosscutting aspects in the area of problem identification and resolution, operating
experience component, because the licensee failed to institutionalize relevant operating
experience in a reasonable time (P.2(b)) (Section 4OA2.1).
TBD. The inspectors identified an apparent violation of Technical Specification 5.8.1.a
(Procedures) for failing to establish a procedure for proper lubrication of the auxiliary
contact sliding mechanism, an activity that affected the performance of the emergency
diesel generator. As a result, craftsmen used an unapproved wet lubricant on the
emergency diesel generator field flash relay auxiliary contact sliding mechanisms, which
was contrary to vendor recommendations, without a procedure that directed the action.
The lubricant was the most likely contributor to oil and dust contamination on the
auxiliary contact surfaces, which apparently caused the emergency diesel generator
failure on February 14, 2007. In addition, a contributor to the apparent violation included
the failure to properly implement the Reliability Centered Maintenance Program.
This finding was greater than minor because it affected the mitigating systems
cornerstone objective (procedure quality attribute), to ensure the availability, reliability,
and capability of systems that respond to initiating events to prevent undesirable
consequences. For the preliminary significance determination, the inspectors used a 14
day exposure time, which was half the time period between the last successful
surveillance and the February 14, 2007, failure. However, this exposure time could
increase to 28 days if the NRC determines the failure was caused by contact binding,
versus contamination. Using the NRC Inspection Manual Chapter 0609, Appendix A,
Determining the Significance of Reactor Inspection Findings for At-Power Situations,
significance determination process, a Region IV senior reactor analyst determined that
the finding was potentially Greater than Green. The finding had crosscutting aspects in
the area of human performance, resources component, in that the licensee failed to
provide a procedure to control a safety related maintenance activity (H.2(c))
(Section 4OA2.2).
Enclosure
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REPORT DETAILS
1.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R19
Post-Maintenance Testing (71111.19)
a.
Inspection Scope
The inspectors selected the below listed risk significant post-maintenance test activity.
The inspectors: (1) reviewed the applicable licensing basis and/or design-basis
documents to determine the safety functions; (2) evaluated the safety functions that may
have been affected by the maintenance activity; and (3) reviewed the test procedure to
ensure it adequately tested the safety function that may have been affected. The
inspectors reviewed test data to verify that acceptance criteria were met.
February 14, 2007, Work Order 263000-02, Replace the Auxiliary Contacts on
the 2CR Starter
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
b.
Findings
Introduction. A self-revealing Green noncited violation of Technical Specification 5.8.1.a
(Procedures) was identified for an inadequate postmaintenance testing procedure.
Craftsmen had replaced the field flash relay auxiliary contacts (following a previous field
flash failure on February 14, 2007) and had misaligned the contact assembly during
installation. Postmaintenance testing was inadequate because it did not verify that the
contacts properly repositioned to the closed position following the surveillance test.
When the emergency diesel generator was started two days later, for a normal
surveillance, the field did not flash because the contacts were stuck open.
Description. When an emergency diesel generator first starts, a control circuit must
energize the generator field flash circuit in order to ensure proper electric generator
operation. A vital component in the field flash circuit is the field flash relay (2CR relay)
and its auxiliary contacts (2CR contacts). The contacts must be in the closed position to
enable the initial field flash but open when the field flash circuit is no longer needed. At
the end of a surveillance run, the contacts must be in their normally closed position to
support flashing the field during the next emergency diesel generator start. If the
contacts fail open, the emergency diesel generator is inoperable.
Following a Train A emergency diesel field flash failure on February 14, 2007, the
licensee found that the field flash relay auxiliary contacts had failed (this failure is
discussed in Section 4OA2 of this report). For this particular failure, high electrical
resistance across the auxiliary contacts likely caused the malfunction. Craftsmen
Enclosure
-5-
replaced the auxiliary contacts and performed postmaintenance testing in accordance
with Work Order 00263000-02. During the postmaintenance test, the emergency diesel
generator started and developed the required voltage. Operators returned the
emergency diesel generator to service. The licensee did not recognize, at the time, that
the field flash relay auxiliary contacts were stuck in the open position or that the
emergency diesel generator was inoperable.
On February 16, 2007, the licensee performed a surveillance of the Train A emergency
diesel generator and the generator field again failed to flash. The licensee found the
stuck open contacts and determined craftsmen had inadvertently misaligned the
contacts during installation, which caused the assembly to bind. The licensees root
cause analysis determined that the postmaintenance test was inadequate because it
failed to verify that the field flash relay auxiliary contacts had returned to their normally
closed position.
Analysis. The failure to perform an adequate postmaintenance test following safety
related work was a performance deficiency. This finding was greater than minor
because the finding was associated with the mitigating systems cornerstone objective
(procedure quality attribute) to ensure the availability, reliability, and capability of
systems that respond to initiating events to prevent undesirable consequences. The
exposure time for this performance deficiency was approximately 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />. Using the
Manual Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection
Findings for At-Power Situations, Phase 1 screening worksheet, the inspectors
determined that the finding was of very low safety significance (Green) because it was
not: 1) a design or qualification deficiency; 2) a loss of system safety function; 3) an
actual loss of safety function for greater than its technical specification allowed outage
time; 4) a loss of safety function of a non-technical specification train; or 5) a seismic,
flooding or severe weather related finding. The finding had crosscutting aspects in the
human performance area, specifically the resource attribute (H.2(c)) in that a complete
and accurate test instruction was not provided to test the 2CR auxiliary relay contacts.
Enforcement. Fort Calhoun Technical Specification 5.8.1.a states, in part, Written
procedures... shall be established, implemented and maintained covering the following
activities... The applicable procedures recommended in Regulatory Guide 1.33,
Revision 2, Appendix A, 1978. Regulatory Guide 1.33, Revision 2, Appendix A, 1978,
Section 8b(1)(q), recommends, in part, specific procedures for surveillance tests,
including "emergency power tests. The licensee performed postmaintenance testing of
the emergency diesel generator in accordance with Work Order 00263000-02. Contrary
to the above, as of February 16, 2007, the field flash relay postmaintenance test
procedure, an emergency power test procedure, was not adequate to satisfy this
requirement because the postmaintenance test failed to verify that the auxiliary contacts
were properly installed. Since this finding was of very low safety significance and was
documented in the licensees corrective action program as Condition
Reports 2007-00745 and 2007-00756, this violation is being treated as a noncited
violation (NCV), consistent with Section VI.A of the NRC Enforcement Policy
(NCV 05000285/2007011-01), Inadequate Emergency Diesel Generator
Postmaintenance Test.
Enclosure
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4.
OTHER ACTIVITIES
4OA2 Identification and Resolution of Problems (71152)
Selected Issue Follow-up Inspection
a.
Inspection Scope
The inspectors selected the below listed issue for an in-depth review. The inspectors
considered the following during the review of the licensee's actions: 1) complete and
accurate identification of the problem in a timely manner; 2) evaluation and disposition
of operability/reportability issues; 3) consideration of extent of condition, generic
implications, common cause, and previous occurrences; 4) classification and
prioritization of the resolution of the problem; 5) identification of root and contributing
causes of the problem; 6) identification of corrective actions; and 7) completion of
corrective actions in a timely manner.
C
February 14, 2007, Train A emergency diesel generator field flash failure
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
b.
Findings
The following two apparent violations are equal contributors to the emergency diesel
generator failure. These apparent violations support one finding of potentially Greater
than Green significance. The NRC will document final enforcement and significance in
future correspondence.
.1
Inadequate Corrective Actions for Emergency Diesel Generator Failure
Introduction. The inspectors identified an apparent violation of 10 CFR 50, Appendix B,
Criterion XVI (Corrective Actions), with two examples, for the failure to: 1) treat the
February 14, 2007, emergency diesel generator failure as a significant condition
adverse to quality; and 2) promptly identify and correct a significant condition adverse to
quality (high resistance on field flash circuit contacts) after determining that similar
operating experience was applicable. In addition, a contributor to the inoperable
emergency diesel generator included the failure to revisit the diesel generator operability
evaluation in response to the applicable operating experience. Overall, the licensee
responded to various problems in isolation and did not adopt a corrective action process
that maintained emergency diesel generator reliability and availability.
Description. When an emergency diesel generator first starts, a control circuit must
energize the generator field flash circuit at 750 rpm in order to ensure proper electric
generator operation. A vital component in the field flash circuit is the field flash relay
(2CR relay) and its auxiliary contacts (2CR contacts). Prior to the 750 rpm engine
speed, the 2CR contacts are in their normally closed position. At 750 rpm the circuit is
Enclosure
-7-
energized and the generator field is flashed. When generator voltage is sufficient, the
generator voltage regulator can control the generator field, thus the field flash circuit is
no longer needed. The field flash circuit is then disabled by energizing the 2CR relay,
which opens the 2CR contacts. The relay is then de-energized and the contacts return
to their normally closed position, ready for the next emergency diesel generator start.
The generator will fail to develop adequate voltage if the field flash circuit does not
function as designed. Two potential failure modes include excessive resistance across
the 2CR contacts (due to surface contamination or oxidation) and mechanical binding of
the contacts in the stuck open position.
On February 14, 2007, during the monthly Train A emergency diesel generator
surveillance, the generator field failed to flash at the required 750 rpm. The licensee
determined that a malfunction of the 2CR auxiliary contacts had caused the field flash
failure.
Inadequate Corrective Actions for February 14, 2007 Failure: On April 30, 2007, the
inspectors identified that the licensee had failed to treat the February 14, 2007,
unsuccessful emergency diesel generator run as a significant condition adverse to
quality and had subsequently failed to take required actions.
A failed emergency diesel generator is a significant condition adverse to quality. For
significant conditions adverse to quality, 10 CFR 50, Appendix B, Criterion XVI
(Corrective Actions) requires the licensee to identify the cause for the failure and to take
actions to preclude repetition. While the licensee did enter the problem into their
corrective action program as Condition Report 200700725 and replaced the 2CR
auxiliary contacts, it failed to identify the cause for the failure or to specify actions to
preclude repetition. This is the first example of an apparent violation fo 10 CFR 50,
Appendix B, Criterion XVI.
Root Cause: In response to the inspectors concerns, the licensee performed a root
cause assessment and sent the failed auxiliary contacts to an independent laboratory
for analysis. The root cause determined that high contact electrical resistance caused
the failure. The licensee had measured contact electrical resistance immediately
afterwards and noted that the as-found resistance varied between 10 and 300 Ohms (on
numerous different measurement attempts). The resistance should have been 0 Ohms.
The licensee determined that as little as 11 Ohms of resistance was sufficient to cause
circuit failure.
An independent laboratory provided a formal report to the licensee entitled Failure
Analysis of GE CR105 X 300 Auxiliary Contact Assemblies, dated July, 2007. The
laboratory found contaminants on the contact surfaces. The contamination was a
combination of oil and dust, which likely came from the operating environment. The
report speculated that some of the oil could have come from the fingers of individuals
who had handled the contacts. However, the report later dismissed this possibility by
stating that some of the contaminants were caked on the surface.
The laboratory cautioned that the reliability of the results was affected by mishandling of
the auxiliary contacts prior to their arrival at the lab. The report stated that the as-found
Enclosure
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condition of the contacts was disturbed and that the auxiliary contacts had been
disassembled and then reassembled in an inappropriate manner. The inspectors
determined that the contacts were mishandled following the diesel failure because the
licensee had not recognized the problems significance and had failed to follow
reasonable protocols for the control and handling of the failed component. Nonetheless,
based on pictures of the contact surfaces, showing spots of contamination, and
electrical resistance readings, the inspectors agreed that the circuit failure was likely
caused by the dust and oil surface contamination. However, neither the inspectors nor
the licensee could conclusively exclude the possibility that the contacts had stuck in the
open position following the previous surveillance (28 days earlier).
Inadequate Response to Applicable Industry Operating Experience: The
inspectors found that the licensee had failed to promptly identify a significant condition
adverse to quality in response to applicable operating experience. Fort Calhoun entered
operating experience associated with an emergency diesel generator failure - dirty field
flash relay contacts - into their Operating Experience System on May 19, 2006 (Fort
Calhoun Operating Experience Number 2007-6767). High resistance across K2 (field
flash interrupt) contacts caused the diesel generator to fail its surveillance test. The K2
contacts serve the same function as the Fort Calhoun 2CR contacts.
The licensee failed to properly evaluate emergency diesel generator operability in
response to the operating experience. Fort Calhoun personnel wrote Condition
Report 200602614. dated June 21, 2006, to evaluate the operating experience. The
operability determination did not challenge operability because, at the time, Operations
did not know if the operating experience applied to Fort Calhoun. Engineering
determined that the operating experience was applicable to Fort Calhoun on September
16, 2006, but did not revisit operability, they just updated the condition report. The
failure to revisit operability when additional information was identified was inconsistent
with NRC RIS 2005-20, Part 9900 Technical Guidance: Operability Determinations and
Functionality Assessments for Resolution of Degraded or Nonconforming Conditions
Adverse to Quality or Safety. Specifically, this document states, in part:
Reviewing the performance of SSCs [structures systems and components] and
ensuring their operability is a continual [emphasis added] process. Potential
degraded or nonconforming conditions of SSCs may be discovered during many
activities:... j. Operational event reviews.
The inspectors determined that the subsequent engineering response and related
corrective actions were not timely. Engineers determined that the operating experience
was applicable to Fort Calhoun on September 16, 2006, but waited until
January 30, 2007, to make a recommendation for a functional test. The due date for
procedure changes was June 15, 2007. The inspectors noted that checking the
resistance across the auxiliary contacts could have been performed in a few minutes.
In addition, the inspectors determined that the licensees response to the subject
operating experience was not reasonable because:
Enclosure
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The operating experience was directly applicable to Fort Calhoun Station.
Fort Calhoun had documented instances of unexpected high resistance readings
on 2CR and 3CR auxiliary contacts. In 1989 and 1990, FCS wrote Work Orders
892492, 900235, and 900485 that documented higher than expected contact
resistance on the Train B emergency diesel generator 2CR and 3CR auxiliary
contacts. The 3CR auxiliary contacts are exactly the same component
(manufacturer and model number) as the 2CR contacts. In response to this
problem, the licensee replaced the affected Train B auxiliary contacts. The Train
A auxiliary contacts were not replaced.
The field flash circuit was particularly sensitive to high resistance across the 2CR
contacts. As little as 11 Ohms could cause field flash circuit failure (similar to the
operating experience description).
The 2CR auxiliary contacts had been in service for many years.
The licensee did not establish a periodic preventive maintenance task to check
contact condition or electrical resistance.
This is the second example of an apparent violation of 10 CFR 50, Appendix B,
Criterion XVI.
Analysis. The failure to treat the February 14, 2007, emergency diesel generator failure
as a significant condition adverse to quality and the failure to promptly identify a
significant condition adverse to quality in response to applicable operating experience
were performance deficiencies. These concerns were greater than minor because they
affected the mitigating systems cornerstone objective (equipment performance
attribute), to ensure the availability, reliability, and capability of systems that respond to
initiating events to prevent undesirable consequences. For the preliminary significance
determination, the inspectors used a 14-day exposure time, which was half the time
period between the last successful surveillance and the February 14, 2007, failure.
However, this exposure time could increase to 28 days if the NRC determines the failure
was caused by contact binding, versus contamination. Using the NRC Inspection
Manual Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection
Findings for At-Power Situations, significance determination process Phase 1 screening
worksheet, the finding screened to a Phase 2 significance determination because the
Train A emergency diesel generator was inoperable for greater than the Technical
Specification allowed outage time. A Region IV senior reactor analyst performed a
Phase 2 significance determination and found the finding was potentially Greater than
Green. The senior reactor analyst performed a preliminary Phase 3 significance
determination, which is included as Attachment 2 to this report. The finding had
crosscutting aspects in the area of problem identification and resolution, operating
experience component, because the licensee failed to institutionalize relevant operating
experience in a reasonable time (P.2(b)).
Enforcement. 10 CFR 50, Appendix B, Criterion XVI (Corrective Actions) states, in part,
Measures shall be established to assure that conditions adverse to quality, such as
failures... [and] deficiencies are promptly identified and corrected. In the case of
Enclosure
-10-
significant conditions adverse to quality, the measures shall assure that the cause of the
condition is determined and corrective action taken to preclude repetition. Contrary to
the above, as of April 30, 2007, the licensee failed to take measures to assure that the
cause of the February 14, 2007, Train A emergency diesel generator failure (a
significant condition adverse to quality) was determined and corrective actions taken to
preclude repetition. In addition, prior to February 14, 2007, the licensee failed to
promptly identify and correct a significant condition adverse to quality, in that the
licensee had determined that operating experience associated with an emergency
diesel generator failure was applicable to Fort Calhoun but failed to promptly identify the
same significant condition at this station. The licensee captured this finding in their
corrective action program as Condition Report 200700725. This is an apparent violation
pending completion of a final significance determination, AV 05000285/20070011-02,
Inadequate Emergency Diesel Generator Corrective Measures.
.2
Failure to Have Procedure for Work on Safety Related Components
Introduction. The inspectors identified an apparent violation of Technical Specification 5.8.1.a (Procedures) for failing to establish a procedure for proper lubrication of the
auxiliary contact sliding mechanism, an activity that affected the performance of the
emergency diesel generator. As a result, craftsmen used an unapproved wet lubricant
on the emergency diesel generator field flash relay auxiliary contact sliding
mechanisms, which was contrary to vendor recommendations, without a procedure that
directed the action. The lubricant was the most likely contributor to oil and dust
contamination on the auxiliary contact surfaces, which apparently caused the
emergency diesel generator failure on February 14, 2007. In addition, a contributor to
the apparent violation included the failure to properly implement the Reliability Centered
Maintenance Program.
Description. The inspectors identified that craftsmen were applying a wet lubricant
(Molykote 55M) to the 2CR auxiliary contact sliding surfaces without an approved
procedure that directed the action. The independent laboratory found a wet lubricant on
the auxiliary contacts sliding surfaces. The licensee had informed the laboratory that
the lubricant was most likely Molykote 55M. This lubricant was the most likely
contributor to oil found on the contact surfaces because oil tends to migrate and the
lubricant was in the closest proximity to the contact surfaces.
The auxiliary contact vendor, General Electric (GE), had recommended against using a
wet lubricant, such as Molykote 55M, because it would attract dirt and foreign material
and could cause auxiliary contact binding. While the licensee had written a letter to GE,
asking for permission to use Molykote 55M, the licensee was unable to find a response
from GE that accepted the practice. More recently, the licensee contacted GE and was
informed that the vendor still did not approve of wet lubricants on the auxiliary contact
sliding mechanisms.
Plant craftsmen informed the inspectors that they always used Molykote 55M on
auxiliary contacts in the plant. The inspectors asked the licensee to provide the
procedure that directed and approved this action. The licensee was unable to provide
an appropriate instruction. The licensee believes that fossil unit workers had started the
informal lubrication practice and permanent plant workers continued it.
Enclosure
-11-
By failing to establish a procedure for this activity, the licensee bypassed other
processes, such as 10 CFR 50.59, that could have identified potential adverse
unintended consequences - such as the increase in the probability of malfunction of
equipment important to safety.
Failure to Follow Reliability Centered Maintenance Program: As a contributing
cause, the licensee determined that their Reliability Centered Maintenance program was
inadequate. The licensee had planned to change to a new program, which would
effectively correct the problem.
The inspectors found that the program was adequate, but the licensee had failed to
properly implement the program. For example, PBD-13, Preventive Maintenance,
Revision 4 states, in part:
The reliability centered maintenance methodology consists of a series of orderly
steps to systematically identify functional subsystems for a system, identify
components required to satisfy the function, and then determine credible failure
modes for each item and the effects that these failure modes would have on
equipment operation.
Preventive Maintenance Basis (PM Basis) - the PM Basis identifies and justified
the PM program on a component [emphasis added] level.
Predictive (Condition Monitoring) Activities - activities that analyze equipment
performance to detect and develop trends so that appropriate corrective actions
can be taken before [emphasis added] the equipment is no longer capable of
performing its intended function...
Preventive Maintenance Standards are developed for each major component
type. Preventive Maintenance Standards include: normal preventive
maintenance to be performed; its frequency and basis; failure mechanisms, their
causes, and preventive maintenance that could be performed on that type of
component; site specific component failure information for that component type;
and recommendations, requirements, and industry operating experience.
Appendix D to PBD-13 specifies, in part, that a comprehensive PM program be
developed because single failures can not be tolerated for risk significant
structures systems and components.
Contrary to the above, for the 2CR relays (with attached 2CR auxiliary contacts):
1.
For the field flash function, the licensee identified the subsystem required to
support the function but failed to determine credible failure modes for each item
or the effects that these failure modes would have on equipment operation.
Consequently, no meaningful preventive maintenance or condition monitoring
was performed.
Enclosure
-12-
2.
No PM Basis was identified and justified on a component [emphasis added] level
basis for the 2CR auxiliary contacts (or the circuit that contained the contacts).
3.
Predictive Maintenance activities were not specified to analyze equipment
performance to detect and develop trends so that appropriate corrective actions
could be taken before [emphasis added] the equipment was no longer capable
of performing its intended function.
4.
A comprehensive preventive maintenance activity was not established for these
risk significant components.
The inspectors noted that the licensee had performed evaluations for other relays and
their associated contacts. The failure modes included dirty contacts. This is a
contributor to the apparent violation. However, failing to establish preventive
maintenance to preclude relay failures meant that the licensee had essentially
implemented a run to failure, then fix approach for these components.
Analysis. The failure to provide a technical specification required procedure to control
the application of lubricants on safety related components was a performance
deficiency. This finding was greater than minor because it affected the mitigating
systems cornerstone objective (procedure quality attribute), to ensure the availability,
reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences. For the preliminary significance determination, the
inspectors used a 14-day exposure time, which was half the time period between the
last successful surveillance and the February 14, 2007, failure. However, this exposure
time could increase to 28 days if the NRC determines the failure was caused by contact
binding, versus contamination. Using the NRC Inspection Manual Chapter 0609,
Appendix AProperty "Inspection Manual Chapter" (as page type) with input value "NRC Inspection Manual 0609,</br></br>Appendix A" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., Determining the Significance of Reactor Inspection Findings for At-Power
Situations, significance determination process Phase 1 screening worksheet, the
finding screened to a Phase 2 significance determination because the Train A
emergency diesel generator was inoperable for greater than the technical specification
allowed outage time. A Region IV senior reactor analyst performed a Phase 2
significance determination and found the finding was potentially Greater than Green.
The senior reactor analyst performed a preliminary Phase 3 significance determination,
which is included as Attachment 2 to this report. The finding had crosscutting aspects
in the area of human performance, resources component, in that the licensee failed to
provide a procedure to control a safety related maintenance activity (H.2(c)).
Enforcement. Fort Calhoun Technical Specification 5.8.1.a states, in part, Written
procedures... shall be established, implemented and maintained covering the following
activities... The applicable procedures recommended in Regulatory Guide 1.33,
Revision 2, Appendix A, 1978. Regulatory Guide 1.33, Revision 2, Appendix A, 1978,
Section 9, recommends procedures for Maintenance that can affect the performance of
safety-related equipment. Contrary to the above, the licensee failed to provide a written
procedure for maintenance that could affect the performance of safety-related auxiliary
contacts, in that craftsmen were applying lubrication to safety-related auxiliary contact
sliding mechanisms without a procedure or other written instruction. The licensee
captured this finding in their corrective action program as Condition Report 2007-2911.
This is an apparent violation pending completion of a final significance determination,
Enclosure
-13-
AV 05000285/20070011-03, Failure to Provide Procedure for Safety Related
Maintenance Activities.
4OA6 Meetings, Including Exit
On September 18, 2007, the Senior Resident Inspector presented the inspection
findings to Mr. Tim Nellenbach, Plant Manager, and other members of the licensees
staff. The licensee acknowledged the findings. The inspector confirmed that proprietary
information was not provided or examined during the inspection.
ATTACHMENTS: 1. SUPPLEMENTAL INFORMATION
2. PRELIMINARY SIGNIFICANCE DETERMINATION
Attachment 1
A1-1
Attachment 1
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
D. Bannister, Acting Site Director
G. Cavanaugh, Supervisor, Regulatory Compliance
M. Frans, Manager, System Engineering
H. Faulhaber, Division Manager, Engineering
M. Ferm, Manager, Shift Operations
D. Guinn, Licensing Engineer
D. Lakin, Manager, Corrective Action Program
J. McManis, Manager, Licensing
T. Nellenbach, Plant Manager
R. Johansan, Manager, Maintenance
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
Inadequate Emergency Diesel Generator
Corrective Measures (Section 4OA2.1)05000285/2007011-03
Failure to Provide Procedure for Safety Related
Maintenance Activities (Section 4OA2.2)
Opened and Closed
Inadequate Emergency Diesel Generator
Postmaintenance Test (Section 1R19)
LIST OF DOCUMENTS REVIEWED
Section 1R19: Post-Maintenance Testing
Work Order 263000-02, Replace the Aux Contacts on the 2CR Starter
Drawing B120C11509, Schematic Diagram Field Flashing Control, Sheet 1, Revision 9
Root Cause Analysis Report, Emergency Diesel Generator #1 (Unknown Inoperability - 2/14/07 to
2/16/07)
Condition Reports 200700756, 200700725, 200700745, and 200700875
Attachment 1
A1-2
Section 4OA2: Other Activities (Identification and Resolution of Problems (71152))
Procedures
SO-R-2, Condition Reporting and Corrective Action, Revision 34
MM-ST-DG-0001, Diesel Generator DG-1 Inspection, Revisions 56 and 57
OP-ST-DG-0001, Diesel Generator 1 Check, Revision 53
PBD-13, Preventive Maintenance, Revision 4
PED-SEI-13, Preventive maintenance Program - Technical Basis, Revision 11
Drawings
B120C11509, Schematic Diagram Field Flashing Control, Sheet 1, Revision 9
B120C11509, Schematic Diagram Field Flashing Control, Sheet 2, Revision 3
44D302335, 1 Phase Full Wave SCR Static Exciter, Sheet 2, Revision 6
B120F11503, emergency Generators Control Cabinets AI-133A & AI-133B, Sheet 3,
Revision 13
Maintenance Work Orders
900485, 892492, 892951, 263000-02
Condition Reports
2007-2227, 2007-2712, 2007-2911, 2007-2912, 200700725, 200700745, 200700756, 200700875
Miscellaneous
Control Room Logs, Day Shift, 01-17-2007
Control Room Logs, Day Shift, 02-14-2007
Control Room Logs, Night Shift, 02-14-2007
Control Room Logs, Day Shift, 02-16-2007
Control Room Logs, Night Shift, 02-16-2007
Shift Manager Logs, Night Shift, 02-16-2007
Southwest Research Institute Failure Analysis of GE CR105 X 300 Auxiliary Contact Assemblies
dated July 2007
Root Cause Analysis Report, Emergency Diesel Generator DG-1 Field Flash Functional Failure
(2/14/2007)
NUREG/CR-5762, Comprehensive Aging Assessment of Circuit Breakers and Relays
TD G080.2800, Instruction Manual for Static Exciter Regulator for AC Generators, Revision 1
TD G100.0490, Instruction Manual for AC Synchronous Generators, Revision 0
Attachment 1
A1-3
TD G080.4350, Renewal Parts Magnetic Motor Starters NEMA Size 3
TD G080.3500, NEMA Size 3 CR105 Magnetic Contactors, CR106 Magnetic Starters
LIST OF ACRONYMS
Apparent Violation
CFR
Code of Federal Regulations
noncited violation
NRC
Nuclear Regulatory Commission
Attachment 2
A2-1
Attachment 2
Preliminary Significance Determination
Significance Determination Basis for Apparent Violations, February 14, 2007 Failure
Reactor Inspection for IE, MS, BI Cornerstones
a.
Phase 1 Screening Logic, Results and Assumptions
In accordance with NRC Inspection Manual Chapter 0612, Appendix B, "Issue Screening,"
the inspectors determined that: 1) the failure to identify the cause and specify corrective
actions to preclude repetition for a significant condition adverse to quality; 2) the
application of a lubricant on the auxiliary contacts without an approved procedure; 3) the
failure to promptly identify and correct a significant condition adverse to quality, that was
identified through operating experience reviews; 4) the failure to revisit operability in
response to new adverse operating experience; and 5) the failure to implement the
Reliability Centered Maintenance program were performance deficiencies. This finding
was determined to be greater than minor because the condition affected the
availability/reliability the Train A emergency diesel generator and thus affected the
Equipment Performance attribute under the Mitigating Systems cornerstone.
The last known successful test of Train A emergency diesel generator occurred 28 days
prior to the February 14, 2007, failure. Because the exact failure mode could not be
determined the exposure time was assumed to be one half the time since the last know
successful performance of the diesel generator or 14 days. In accordance with NRC
Inspection Manual Chapter 0609, Appendix A, Determining the Significance of Reactor
Inspection Findings for At-Power Situations, dated March 23, 2007, the inspectors
conducted a significance determination process (SDP) Phase 1 screening and determined
that the finding resulted in loss of the safety function of Train A emergency diesel
generator for greater than the Technical Specification allowed outage time. Consequently,
a Phase 2 SDP risk significance estimation was required.
b.
Phase 2 Estimation
Internal Events and Large Early Release Frequency (LERF)
The inspectors and a RIV senior reactor analyst (SRA) performed a Phase 2 evaluation
using the Risk-Informed Inspection Notebook for Fort Calhoun Power Station,
Revision 2.01, (SDP Phase 2 Notebook) and its associated "Phase 2 Pre-solved Table."
The last successful surveillance test of Train A emergency diesel generator was performed
on January 17, 2007, and the test failure was on February 14, 2007. The inspectors could
not determine with certainty when the EDG became non-functional. Therefore, in
accordance with the SDP Usage Rule for exposure time, a "T/2" exposure time of 14 days
was assumed. A 14-day exposure time is represented as a "3 - 30 day" exposure category
in the SDP Phase 2 evaluation.
Attachment 2
A2-2
The inspectors determined that after an emergency start of Train A emergency diesel
generator during a loss of offsite power (LOOP), operators would not be capable of
diagnosing and correcting the problem with the field flash relay. Therefore, recovery was
not credited.
The Train A emergency diesel generator was identified as a target in the Phase 2 pre-
solved table so it could be used directly to assess the finding. The pre-solved table
identified that the finding was "CDF-dominant." Therefore, no additional review was
required for LERF consideration. For a 3 - 30 day exposure time, the pre-solved table
identified that the significance of the finding was White with respect to CDF. The dominant
sequence (with an equivalent risk contribution of 6) involved a station blackout (LOOP with
failure of the EDGs), and failure to recover offsite power in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (LOOP - EAC - REC8).
External Events
Neither the Fort Calhoun SDP Phase 2 Notebook, nor the pre-solved table includes
screening capability for external events or other initiating events. Because the risk
contribution of the finding due to internal events was Greater than Green, the SRA
conducted additional review for external event contribution. Experience has shown, using
the Risk-Informed Inspection Notebooks, that accounting for external initiators could result
in increasing the risk significance of an inspection finding by as much as one order of
magnitude. The SRA determined that the most efficient method of accounting for external
initiators was to use the guidance provided in IMC 0609, Appendix A, Attachment 3, "User
Guidance for Screening of External Events Risk Contributions."
Screening of Fire Risk Contributions
The finding affected Train A emergency diesel generator, which was not in the protected
train of the post-fire safe shutdown path. Ordinarily, this would result in the finding being
screened from further consideration for fire-risk contribution using IMC 0609, Appendix A,
Attachment 3. However, the licensee's probabilistic safety assessment staff informed the
analyst that there were some fire areas where Train A emergency diesel generator was
credited. These fire areas included fires in the upper electrical penetration room, the lower
electrical penetration room, and for any fire scenario involving a LOOP with the loss of the
4160 V Bus 1A4 or the EDG-2. The risk contribution of these fire areas is addressed in the
Section C.
Screening of Seismic Risk Contributions
The analyst determined the finding did not immediately screen out as insignificant due to
seismic risk contribution. This was because Train A emergency diesel generator was on
the licensee's IPEEE list of equipment addressed in their seismic margins approach to
seismic risk contribution, it was used to mitigate the consequences of a loss of AC power
during a seismic event, and the exposure time of the finding was greater than 3 days.
Therefore, the analyst continued the evaluation using the guidance in the Risk Assessment
Standardization Project (RASP) External Events Handbook.
The analyst estimated the seismic contribution using the "Seismic Event Modeling and
Seismic Risk Quantification Handbook" of the RASP External Events Handbook. For Fort
Attachment 2
A2-3
Calhoun Station, the frequency of a seismically-induced LOOP event was estimated as
1.06E-4/year. The SPAR model estimate of the failure-to-run of EDG-2 was 2.068E-2.
Therefore the estimated CDF of a seismically-induced LOOP with a random failure of
DG-2 for a 14-day exposure period was in the high E-8/year range. The seismic risk
contribution of the finding was insignificant relative to the internal events result.
Screening of Flood Risk Contributors
Using IMC 0609, Appendix A, Attachment 3, Table 3.1, "Plant Specific Flood Scenarios
and Initiator Frequencies," the analyst determined that Train A emergency diesel generator
was not a structure, system, or component identified as critical to avoiding core damage for
any flood scenario of significance. Therefore, flood risk contribution was screened out
from further consideration.
Estimation of External Event Risk Contributions
Based on the screening of external event risk contributions described above, the analyst
concluded the total risk contribution from external events (other than fire) resulting from the
finding was not significant. Further assessment was necessary with respect to fire-risk
contribution.
c.
Validation of Phase 2 SDP Results
The SRA used the NRC's simplified plant analysis risk (SPAR) model for Fort Calhoun
Station, Revision 3.31, dated April 10, 2006, to estimate the increase in risk associated
with the finding due to internal initiating events. Average test and maintenance and
"winter" raw water success criteria were assumed. A cutset truncation of 1.0E-12 was
used.
Consistent with the Phase 2 SDP evaluation, the SRA used a 14-day exposure time and
the observed failure of Train A emergency diesel generator was modeled as a failure-to-
start basic event.
The SRA noted that an influential assumption in the analysis would be whether the finding
resulted in a random, independent, or common-cause failure of the Train A emergency
diesel generator field flash relay. Guidance in the RASP Handbook states that if the cause
of the failure cannot be determined to be independent, or common-cause related, that the
cause should be assumed to be random in the analysis. Additionally, the RASP Handbook
states that a component failure can be considered independent when the cause is well
understood and there is no possibility (probability = 0.0) that the same circumstances exist
in other components in the common-cause component group.
At the conclusion of the inspection, the licensee had not definitively ruled out the potential
for common-cause failure to be a consideration in its root cause assessment work. Some
evidence indicated that the Train A emergency diesel generator field flash relay failure may
have been independent based on differences in the age of the relays and past
maintenance practices. However, the SRA considered that there must be at least some
possibility that the same circumstances resulting in the Train A emergency diesel generator
failure could exist in EDG-2.
Attachment 2
A2-4
Consistent with guidance in the RASP Handbook, including the NRC document, "Common-
Cause Failure Analysis in Event Assessment, (June 2007)", the SRA modeled the
condition by adjusting the following basic events in the SPAR model:
EPS-DGN-FS-1A = TRUE
EPS-DGN-FR-1A = 1.0
EPS-DGN-CF-RUN = FALSE
The SPAR baseline CDF was 1.69E-5/year. The evaluation case for the above change set
resulted in a CDF of 2.40E-4/year. The dominant core damage sequence was a LOOP,
followed by failure EDG-2, failure to maintain reactor coolant system subcooling, and
failure to recover an EDG or offsite power within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
CDF = CDFcase - CDFbase
= 2.40E-4/year - 1.69E-5/year = 2.23E-4/year.
Therefore, the total change in core damage frequency over the exposure time that was
related to this finding was calculated as:
CDF = 2.23E-4/year ÷ 365 days/year * 14 days = 8.56E-6 over the period.
This result was consistent with the White Phase 2 SDP result. Therefore, the analyst
considered the White Phase 2 SDP result to be validated for the risk contribution resulting
from internal initiating events.
Potential Risk Contribution due to LERF,
The dominant core damage sequences evaluated in the Phase 2 SDP evaluation did not
involve a potential for containment bypass. Therefore, using the guidance of IMC 0612,
Appendix HProperty "Inspection Manual Chapter" (as page type) with input value "NRC Inspection Manual 0612,</br></br>Appendix H" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., "Containment Integrity Significance Determination Process," the finding
screened from further consideration with respect to LERF. The analyst concluded that the
significance of the finding obtained from the CDF assessment was sufficient to
characterize its risk significance and no further LERF-related evaluation was necessary.
Estimation of Risk Contribution from Fire and Total Risk Contribution of Finding
Using information provided by the licensee from its own fire analysis work, the analyst
performed a simplified estimate of the total risk contribution (including fire events) using the
internal events SPAR model result. The licensee estimated that for a finding resulting in
unavailability of the Train A emergency diesel generator, approximately 33 percent of the
total core damage frequency resulted from fire events.
Therefore, for a 14-day exposure period, the total risk contribution of the finding for internal
initiating events and external initiating events combined could result in a CDF in the low
E-5 range.
Attachment 2
A2-5
d.
Licensees Preliminary Risk Evaluation
The licensee's probabilistic safety assessment (PSA) staff provided the analyst with
preliminary insights from its own risk assessment of the Train A emergency diesel
generator failure. The licensee first performed its assessment assuming Train A
emergency diesel generator was out of service for maintenance, which did not address the
concern discussed above addressing adjustments for potential common-cause failure. The
licensee's baseline model CDF, which also includes seismic initiators, was 1.87E-5/year.
The evaluation case CDF with Train A emergency diesel generator in maintenance was
7.54E-5/year, resulting in a CDF of 2.2E-6/year over the 14-day exposure period.
The licensee provided additional information after consideration of fire scenarios and
consideration of common cause failure to start of EDG-2. The new model added the
existing licensees existing Revision 7 PRA model to an upgraded fire model taken from
the Individual Plant Examination of External Events (IPEEE). An example of an upgrade in
the fire model obtained from the IPEEE was that the probability of circuit failure resulting
from a hot short was increased from the IPEEE original value of 0.06 to a revised value of
0.6.
In the revised model, which now included internal events, internal flood, seismic, and fire,
the baseline CDF with random failures of the EDGs and nominal common cause failure,
was 4.08E-5/year. With Train A emergency diesel generator failed, and a change in the
probability of EDG-2 failure due to common cause increasing similar to the SPAR model
calculation, the resulting evaluation case CDF was 1.75E-4/year. This resulted in a CDF
of 5.14E-6/year over the 14-day exposure period.
The dominant fire initiating event was a fire in the EDG-2 room, with subsequent failure of
EDG-2, hot short-induced failure of offsite power to its associated vital bus, and failure of
offsite power to the other train with the already failed Train A emergency diesel generator.
Core damage sequences involving this initiator represented 13 percent of the total
contribution to CDF.
e.
Conclusion
The Phase 2 analysis of the significance of the finding, validated with use of the NRC
SPAR model, resulted in a determination that the significance was at least of low-to-
moderate safety significance (White). However, this conclusion was dependent on the
influential assumption of a 14-day exposure time. Additionally, external initiating events
were known to be a potentially significant contributor to the overall significance of the
finding. Although the licensees preliminary results had been discussed, the licensee had
not completed its own final analysis of the significance of the finding by the conclusion of
the inspection. Pending completion of this analysis and review by the NRC staff, the
analyst considered the significance of the inspection finding best characterized as Greater
than Green.