ML112000064
| ML112000064 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 07/18/2011 |
| From: | Collins E Region 4 Administrator |
| To: | Bannister D Omaha Public Power District |
| References | |
| EA-11-025 IR-11-007 | |
| Download: ML112000064 (11) | |
See also: IR 05000285/2011007
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGI ON I V
612 EAST LAMAR BLVD, SUITE 400
ARLINGTON, TEXAS 76011-4125
July 18, 2011
David J. Bannister, Vice President
and Chief Nuclear Officer
Omaha Public Power District
Fort Calhoun Station FC-2-4
P.O. Box 550
Fort Calhoun, NE 68023-0550
SUBJECT: FORT CALHOUN STATION - FINAL SIGNIFICANCE DETERMINATION
FOR A WHITE FINDING AND NOTICE OF VIOLATION, NRC INSPECTION
REPORT 05000285/2011007
Dear Mr. Bannister:
The purpose of this letter is to provide you the final significance determination of the preliminary
Yellow finding identified in our previous communication dated May 6, 2011, which included the
subject inspection report. The inspection finding was assessed using the Significance
Determination Process and was preliminarily characterized as a Yellow finding with substantial
importance to safety that may result in additional NRC inspection and potentially other NRC
action. This finding was associated with the June 14, 2010, failure of a reactor trip
contactor (M2) in your reactor protection system.
At your request, a regulatory conference was held on June 2, 2011, to further discuss your
views on this issue. During the regulatory conference, your staff described the Fort Calhoun
Stations assessment of the significance of the finding and they provided a summary of the
corrective actions, and insights from the root cause analysis of the finding. This material is
documented in the NRC Meeting Summary (ML111660027) dated June 14, 2011. You also
requested that the NRC reconsider its evaluation of the findings risk significance based on four
specific areas of consideration where differences exist between the NRCs preliminary
significance determination and your staffs risk assessment. These are: 1) Shorter Exposure
Time (T/2 + repair vs. T + repair); 2) Lower Failure Probability for Clutch Power Supply Breaker;
3) Common Cause Failure Determination; and 4) Higher Operator Reliability in Tripping the
Reactor. Between June 6 and June 28, 2011, you provided supplemental information regarding
follow-up questions asked by NRC staff at the conference. This additional material was
docketed as ADAMS document ML111881131.
The NRC has reviewed your areas of consideration and our evaluation of each is provided in
Enclosure 2 of this letter along with the revised NRC risk assessment. The NRC considered the
information developed during the inspection, and the information that you provided at, and
subsequent to, the conference. The NRC has concluded that the finding is appropriately
Omaha Public Power District -2- EA-11-025
characterized as White, a finding with low to moderate importance to safety and will result in
additional NRC inspection and potentially other NRC actions.
You have 30 calendar days from the date of this letter to appeal the staffs determination of
significance for the identified White finding. Such appeals will be considered to have merit only
if they meet the criteria given in NRC Inspection Manual Chapter 0609, Attachment 2. An
appeal must be sent in writing to the Regional Administrator, U.S. Nuclear Regulatory
Commission, Region IV, 612 E. Lamar Blvd., Suite 400, Arlington, Texas 76011-4125.
The NRC has concluded that failure to assure that the cause of a significant condition adverse
to quality was determined and failure to take corrective actions to preclude repetition of the
condition, is a violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50,
Appendix B, Criterion XVI, Corrective Action, as cited in the enclosed Notice of Violation. The
circumstances surrounding the violation are described in detail in the subject inspection report.
In accordance with the NRC Enforcement Policy, the Notice of Violation is considered an
escalated enforcement action because it is associated with a White finding.
You are required to respond to this letter. Please follow the instructions specified in the
enclosed Notice of Violation when preparing your response. If you have additional information
that you believe the NRC should consider, you may provide it in your response to the Notice.
The NRC review of your response to the Notice will also determine whether further enforcement
action is necessary to ensure compliance with regulatory requirements.
Because your current plant performance is in the Degraded Cornerstone (Mitigating Systems)
Column, and this violation also impacts that cornerstone, the NRC will use the NRC Action
Matrix to determine the most appropriate NRC response to this violation. The NRC will notify
you, by separate correspondence, of that determination.
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its
enclosures, and your response will be available electronically for public inspection in the NRC
Public Document Room or from the NRCs document system (ADAMS). ADAMS is accessible
from the NRC Web site at www.nrc.gov/reading-rm/adams.html. To the extent possible, your
response should not include any personal privacy, proprietary, or safeguards information so that
it can be made available to the Public without redaction.
Sincerely,
/RA/
Elmo E. Collins
Regional Administrator
Docket: 50-285
License: DPR-40
Enclosures:
1. Notice of Violation
Omaha Public Power District -3- EA-11-025
2. Fort Calhoun Reactor Protection System Issue
Final Significance Determination
cc w/Enclosures:
Distribution via Listserv
Omaha Public Power District -4- EA-11-025
Electronic distribution by RIV:
Regional Administrator (Elmo.Collins@nrc.gov)
Deputy Regional Administrator (Art.Howell@nrc.gov)
DRP Director (Kriss.Kennedy@nrc.gov)
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DRS Director (Anton.Vegel@nrc.gov)
Acting DRS Deputy Director (Robert.Caldwell@nrc.gov)
Senior Resident Inspector (John.Kirkland@nrc.gov)
Resident Inspector (Jacob.Wingebach@nrc.gov)
Acting Branch Chief, DRP/E (Ray.Azua@nrc.gov)
Senior Project Engineer, DRP/E (Ray.Azua@nrc.gov)
Project Engineer (Jim.Melfi@nrc.gov)
Project Engineer (Chris.Smith@nrc.gov)
RIV Enforcement, ACES (Ray.Kellar@nrc.gov)
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Project Manager (Lynnea.Wilkins@nrc.gov)
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OEMail Resource
DRS/TSB STA (Dale.Powers@nrc.gov)
RIV/ETA: OEDO (John.McHale@nrc.gov)
R:_\Reactors\FCS\FCS-Final-Significance.docx
ADAMS Yes SUNSI Review Complete Reviewer Initials: JAC
Publicly Available Non-publicly Available Sensitive Non-sensitive
RIV/DRP:PBE DRP:PBE DRS-SRA D:DRS ACES
RVAzua JAClark DPLoveless AVegel RKellar
/RA/ /RA/ /RA/ /RA/ /RA/via email
07/08/11 07/08/11 07/14/11 07/14/11 07/07/11
Counsel NRR/OE D:DRP ORA
MBarkman Marsh NColeman KMKennedy EECollins
/RA/via email /RA/via email /RA/ /RA/
07/13/11 07/13/11 07/15/11 07/18/11
OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
NOTICE OF VIOLATION
Omaha Public Power District Docket No.: 05000285
Fort Calhoun Station License No.: DPR-40
During an NRC inspection conducted from January 17 through April 15, 2011, one violation of
NRC requirements was identified. In accordance with the NRC Enforcement Policy, the
violation is listed below:
Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI,
Corrective Action, requires, in part, that measures shall be established to assure that
conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations,
defective material and equipment, and nonconformances are promptly identified and
corrected. In the case of 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, between November 3, 2008, and June 14, 2010, the licensee
failed to assure that the cause of a significant condition adverse to quality was
determined and corrective actions were taken to preclude repetition. Specifically, the
licensee failed to preclude shading coils from repetitively becoming loose material in the
M2 reactor trip contactor. The licensee failed to identify that the loose parts in the trip
contactor represented a potential failure of the contactor if they became an obstruction;
and therefore, failed to preclude repetition of this significant condition adverse to quality,
that subsequently resulted in the contactor failing.
This violation is associated with a White significance determination process finding in the
Mitigating Systems Cornerstone.
Pursuant to the provisions of 10 CFR 2.201, Omaha Public Power District is hereby required to
submit a written statement or explanation 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 IV, 612 E. Lamar Blvd, Suite 400,
Arlington, Texas, 76011-4125, and a copy to the NRC Resident Inspector - Fort Calhoun
Station, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This
reply should be clearly marked as a "Reply to a Notice of Violation; EA-11-025" and should
include for each violation: (1) the reason for the violation, or, if contested, the basis for disputing
the violation or severity level, (2) the corrective steps that have been taken and the results
achieved, (3) the corrective steps that will be taken, and (4) the date when full compliance will
be achieved. Your response may reference or include previous docketed correspondence, if
the correspondence adequately addresses the required response. If an adequate reply is not
received within the time specified in this Notice, an order or a Demand for Information may be
issued as to why the license should not be modified, suspended, or revoked, or why such other
action as may be proper should not be taken. Where good cause is shown, consideration will
be given to extending the response time.
-1- Enclosure 1
If you contest this enforcement action, you should also provide a copy of your response, with
the basis for your denial, to the Director, Office of Enforcement, United States Nuclear
Regulatory Commission, Washington DC 20555-0001.
Because your response will be made available electronically for public inspection in the NRC
Public Document Room or from the NRCs document system (ADAMS), accessible from the
NRCs website at www.nrc.gov/reading-rm/adams.html, to the extent possible, it should not
include any personal privacy, proprietary, or safeguards information so that it can be made
available to the public without redaction. If personal privacy or proprietary information is
necessary to provide an acceptable response, then please provide a bracketed copy of your
response that identifies the information that should be protected and a redacted copy of your
response that deletes such information. If you request withholding of such material, you must
specifically identify the portions of your response that you seek to have withheld and provide in
detail the bases for your claim of withholding (e.g., explain why the disclosure of information will
create an unwarranted invasion of personal privacy or provide the information required by
10 CFR 2.390(b) to support a request for withholding confidential commercial or financial
information). If safeguards information is necessary to provide an acceptable response, please
provide the level of protection described in 10 CFR 73.21.
In accordance with 10 CFR 19.11, you may be required to post this Notice within two working
days.
Dated this 18th day of July 2011
-2- Enclosure 1
Fort Calhoun Station Reactor Protection System Issue
Final Significance Determination
During the regulatory conference held on June 2, 2011, the Fort Calhoun Station (FCS) staff
described your assessment of the significance of the finding as summarized below. Specifically,
your staff discussed four differences that existed between the NRCs preliminary significance
determination and your risk assessment. These differences and our conclusions are as follows:
Item 1 - Shorter Exposure Time (T/2 + repair vs. T + repair)
Your staff stated that exposure time for this issue should not utilize T plus repair time, but use
T/2 plus repair time instead. This would result in a reduced exposure period from 64.0 days to
32.5 days. This was based on your analysis that a shading coil must fragment, due to wear, prior
to a piece of it being able to jam the contactor in the closed position. You also stated this wear
would likely take weeks or months. Therefore, you concluded that the fragmenting and jamming
occurred at some unknown time between April 10, and June 14, 2010. This would indicate that
the use of T/2 is more applicable to this case.
NRC staff determined that the provided failure modes and effects analysis for the shading coil
was very comprehensive and understandable. However, there was no corresponding failure
modes and effects analysis presented for the overall contactor (i.e., how the shading coil failure
could cause the contactor failure). Definitive testing or evaluation of the jamming sequence for
the contactor was not provided.
During discussions with your forensic specialist at the regulatory conference, NRC staff
questioned the methods used to determine how the shading coil actually jammed the contactor.
The specialist indicated that specific confirmation testing was not conducted, but that a shading
coil fragment was likely repositioned during vibration, moved in an upward direction, and then
jammed the contactor mechanism in its opening motion on June 14, 2010. Based on visual and
physical evidence, NRC staff concluded that this was unlikely. The travel on the contactor
mechanism, from full contact closure until the contacts open, was only approximately 1/8 inch.
The NRC staff concluded it would be extremely difficult for a shading coil fragment to both enter
the gap between the frame and the contactor slide and stop the contactor slide from moving in
such a small amount of travel. However, when a contactor slide moves from the full open to the
closed position, the travel is over 1/2 inch. The NRC staff believes it is more likely a whole
shading coil or fragment was forced into the gap between the frame and the contactor slide
during a closing action; specifically the April 10, 2010, closing prior to the June 14, 2010, failure.
Therefore, the NRC concludes the applicable exposure time was 63 days, plus a 1 day repair
time, for a total of 64 days.
Item 2 - Lower Failure Probability for Clutch Power Supply Breaker
Your staff stated that the generic breaker failure data used in the preliminary significance
determination was not the best available information for vital breakers CB-AB and CB-CD.
Instead your staff suggested that the NRC staff use generic data from NUREG/CR-6928,
Industry-Average Performance for Components and Initiating Events at U.S. Commercial
Nuclear Power Plants, plus data developed using test results from testing the two breakers
previously installed at Fort Calhoun. However, your final assessment indicated that you believed
a Bayesian update of the test data, using a Jeffreys non-informative prior distribution would be
the appropriate value.
The NRC staff determined that, to the extent the test data from the previously installed breakers
represented the installed conditions of the breakers, this data should be used to update the
generic data. However, the NRC staff concluded that the test data should not be used to update
-1- Enclosure 2
Fort Calhoun Station Reactor Protection System Issue
Final Significance Determination
a Jeffreys non-informative prior distribution when existing generic priors were available that
adequately represented the population of the breakers in question. The staff also concluded that
data from NUREG/CR-6928 should not be used because the breakers in question were neither
reactor trip breakers nor were they maintained and tested to the standards used for reactor trip
breakers.
The NRC staff updated the priors used in the preliminary significance determination with the data
obtained from the test results on vital breakers CB-AB and CB-CD. The NRC concluded that this
approach represented the best available information. The calculated total failure probability for
the breakers was 3.81 x 10-4 demand which is a change from 7.5 x 10-3 documented in the
preliminary determination.
Item 3 - Common Cause Failure Determination
Your staff stated that there was no single clear path for analysis of common cause failure for this
issue and recommended that the NRC staff use the definition of common cause failure
documented in NUREG/CR-5500, Volume 10, Reliability Study: Combustion Engineering
Reactor Protection System, 1984-1998. Additionally, your staff commented that the NRC staff
made an incorrect reference to Revision 1.01 of the Risk Assessment of Operational Events
handbook in our inspection report. Finally, your staff stated that the common cause observations
in the inspection report under Assumption 7 may need to be updated based on new information
provided in the Engineering Systems, Inc. report.
The NRC staff determined that the reference to Revision 1.01 of the handbook was incorrect.
However, this definition was not used in the common cause methodology utilized in our analysis.
The reasons for adjusting the common cause failure probability were best described in the
inspection report Page A-4, Assumptions 7 and 8.
The NRC staff also determined that NUREG/CR-5500 provides a concise definition of a common
cause failure. However, in the significance determination, the NRC staff did not assume that a
common cause failure event had occurred. If a failure of Contactors M1 and M2 had occurred at
the same time, the risk would have been significantly higher than our original estimates. The
guidance contained in NUREG/CR-5500 was not intended to be used to evaluate a condition
where the analyst believes that the common cause failure probability should be increased based
on observed conditions. The NRC staff has determined that the approach used in the inspection
report is the appropriate method to adjust common cause failure probabilities when components
are maintained and operated under similar conditions.
The NRC staff reviewed Assumption 7 in the NRC inspection report in light of the findings
documented in the report generated by the professional engineering consulting firm Engineering
Systems, Inc. However, the only condition that may have changed based on the Engineering
Systems, Inc. report was that, subparts exhibited significant scratching and indentations. The
NRC staff determined that despite such a change, the subject conditions, operation and
maintenance history of the contactors still warranted adjustment of the common cause failure
probability of contactor M1 given that contactor M2 failed.
Common cause failure probabilities are included in probabilistic risk assessment because
analysts have long recognized that many factors, such as the poor maintenance practices
indicated in the inspection report, which are not modeled explicitly in the models, can defeat
redundancy or diversity and make failures of multiple similar components more likely than would
be the case if these factors were absent. The effect of these factors on risk can be significant.
-2- Enclosure 2
Fort Calhoun Station Reactor Protection System Issue
Final Significance Determination
For practical reasons related to data availability, the common cause failure probabilities of similar
components are estimated using data collected at the component level, without regard to failure
cause.
Factors such as poor maintenance processes are often part of the environment in which the
components are embedded and are not intrinsic properties of the components themselves. The
NRC staff uses the failure memory approach in evaluating the significance of a performance
deficiency. Observed failures are mapped into the probabilistic model, but successes are treated
probabilistically. Thus, failure probabilities are left at their nominal values or are conditioned as
necessary to reflect the details of the event.
To address this conditioning, the NRC staff has determined that there are three basic ground
rules for treatment of common cause failure:
a. The shared cause is the deficiency identified in the inspection report which led to the
observed equipment failure. In the case of the subject finding, the licensees failure to
identify the cause of the loose shading coils was the performance deficiency. The
inspectors observed that at least one shading coil would easily come out of its recess on
all contactors.
b. Common cause failures are of concern when they occur during the mission time of the
probabilistic risk assessment, which for internal hazard groups is generally 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The
common cause failure analysis methodology used and alpha vectors documented in the
inspection report were developed to intrinsically incorporate this requirement into the
common cause failure probabilities.
c. Credit for programmatic actions to mitigate common cause failure potential (staggering
equipment modifications, etc.) should be applied qualitatively during the enforcement
process and not incorporated into the numerical risk result. For the subject performance
deficiency, this condition is moot. Inspection of components and records reviews
indicated that all contactors had been handled in the same manner.
Therefore, the NRC concludes that the treatment of common cause failure probabilities for the
reactor protection system contactors was appropriate and the conditional failure probability of the
M1 contactor is best approximated as 3.59 x 10-2/demand.
Item 4 - Higher Operator Reliability in Tripping the Reactor
Item 4a - Under Anticipated Transient Without Scram Conditions
Your staff indicated that follow-up operator actions, past the 10-minute point in the anticipated
transient without scram (ATWS) scenario, should be credited. You provided an evaluation by
Westinghouse of the expected Fort Calhoun Station plant response to this event. The evaluation
indicated that, due to a large negative moderator temperature coefficient, power would
automatically be reduced before the American Society of Mechanical Engineers (ASME) Level C
pressure limit of 3200 psig was exceeded. This would indicate that further operator actions could
be taken to trip the control rods without physical damage to key reactor components or systems.
NRC staff determined that the reactor response to a delayed tripping of the control rods in an
ATWS scenario, especially the pressure response, is a critical aspect in preventing core damage.
The details of the calculations and thermal-hydraulic runs of record are well established.
NUREG-1780 states that pressure transients are unacceptable if the ASME Level C value of
3200 psig is exceeded. It further stated that a higher ASME service level was considered for
-3- Enclosure 2
Fort Calhoun Station Reactor Protection System Issue
Final Significance Determination
Babcock & Wilcox and Combustion Engineering plants, but was rejected on the basis that the
reactor coolant system pressure boundary could deform to the point of inoperability.
Your evaluation showed a peak pressure of 3176 psia (approximately 3162 psig) during a run of
the Combustion Engineering Nuclear Transient Simulator (CENTS) code. The NRC noted that
similar thermal-hydraulic code runs, referenced in NUREG-1000 and NUREG-1780, were very
sensitive to small variations or uncertainties in plant-specific parameters such as moderator
temperature coefficient, reactor vessel volumes, and other physical parameters. Your analysis
did not include sensitivities to variations or uncertainties in these parameters. For example, your
analysis used the Fort Calhoun Station predicted beginning of life full power moderator
temperature coefficient. However, you did not provide a sensitivity analysis for moderator
temperature coefficient showing potential inaccuracies in this value or its variation with power.
NUREG-1780 states that during the first part of the fuel cycle, below 100 percent power, the
moderator temperature coefficient can be positive or insufficiently negative. If an ATWS occurs
when the moderator temperature coefficient is either positive or insufficiently negative to limit
reactor power, and the ATWS pressure increases, all subsequent mitigating functions are likely to
be ineffective. NRC staff reviewed your predicted moderator temperature coefficient values over
core life and at different power levels and concluded you also have positive or insufficiently
negative values at lower powers.
It is the NRCs judgment that the 3176 psia outcome of your analysis is insufficient to assure the
ASME Level C value is not actually exceeded, considering the potential inaccuracies and
uncertainties of the analysis. Therefore, the NRC concluded the preliminary assessment time
limitations for the ATWS response should still be used and no changes were made to the
assessment for additional operator actions beyond 10 minutes.
Item 4b - Manual Trip Probability
Your staff pointed out that the failure of operators to push manual trip pushbutton No. 2 was not
dependant on the success or failure of manual trip pushbutton No. 1. Based on your procedures
the NRC staff concluded that, based on procedural guidance and operator training, the failure of
operators to push manual trip pushbutton No. 2 would not likely be affected by the success or
failure of manual trip pushbutton No. 1. Therefore, additional credit was given for the former
probability under RPS-XHE-ERROR as shown in Table 1. However, the NRC did not use your
suggested values (6 x 10-4) for either manual pushbutton, as those values were based on
additional time available to the operators in an ATWS scenario which the NRC staff determined
should not be credited as discussed in Item 4a.
-4- Enclosure 2
Fort Calhoun Station Reactor Protection System Issue
Final Significance Determination
Summary
Table 1
Summary of Parameter Changes
Fort Calhoun Station Reactor Protector System Contactor Issue
Final Significance Determination
Parameter Basic Event SPAR Preliminary Licensee Final
Value Significance Recommended Significance
1 Shorter Exposure Time N/A N/A 64 days 32.5 days 64 days
-3 -3 -4 -4
2 Lower Failure Probability for RPS-BSN-FO-CBAB 7.5 x 10 7.5 x 10 1.2 x 10 3.81 x 10
Clutch Power Supply Breaker RPS-BSN-FO-CBCD
-6 -2 -6 -2
3 Common Cause Failure RPS-RYT-CF-M12 2.4 x 10 3.59 x 10 2.4 x 10 3.59 x 10
-4
3 Contactor Failure RPS-RYT-CC-M1 1.2 x 10 1.0 1.0 1.0
-3
4a Operator Reliability Under N/A N/A N/A 1.4 x 10 N/A
-2 -3 -4 -3
4b Manual Trip 1 RPS-XHE-XM- 1 x 10 1.5 x 10 6.0 x 10 1.5 x 10
-4 -3
4b Manual Trip 2 RPS-XHE-ERROR N/A 0.5 6.0 x 10 6.0 x 10
The NRC staff requantified the detailed model of the reactor protection system used in the
preliminary significance determination using the modified parameters listed in Table 1. The
revised internal change in core damage frequency was calculated to be 6.47 x 10-6. Combining
this with the external risk calculated in the preliminary determination the total change in core
damage frequency was 7.14 x 10-6.
The staff has considered the information you provided to the NRC regarding the significance of
this issue and has concluded that the finding is appropriately characterized as being of low to
moderate safety significance (White). The agencys preliminary evaluation, as documented in
NRC Inspection Report 05000285/2011007, has been modified as shown above to reflect that the
change in core damage frequency for the finding was 7.14 x 10-6 as compared with 2.6 x 10-5.
-5- Enclosure 2