ML050490461
| ML050490461 | |
| Person / Time | |
|---|---|
| Site: | Kewaunee  |
| Issue date: | 02/18/2005 |
| From: | Satorius M Division of Nuclear Materials Safety III |
| To: | Lambert C Nuclear Management Co |
| References | |
| IR-04-009 EA-05-021 | |
| Download: ML050490461 (5) | |
See also: IR 05000305/2004009
Text
February 18, 2005
EA 05-021
Mr. Craig Lambert
Site Vice President
Kewaunee Nuclear Power Plant
Nuclear Management Company, LLC
N490 State Highway 42
Kewaunee, WI 54216-9511
SUBJECT: PRELIMINARY SIGNIFICANCE DETERMINATION FOR A GREATER THAN
GREEN FINDING (NRC INSPECTION REPORT 50-305/2004-09) - KEWAUNEE
CONTAINMENT EQUIPMENT HATCH INTERFERENCE
Dear Mr. Lambert:
The purpose of this letter is to provide you with the Nuclear Regulatory Commissions (NRC's)
preliminary significance determination for the performance deficiency which was described in
NRC Inspection Report 50-305/2004-09 and involved the inability of your staff to rapidly close
the containment equipment hatch during cold shutdown conditions due to an interference. The
interference was caused by the inadequate design of a rail system which was installed in the
containment to facilitate reactor vessel head replacement activities. The preliminary
significance determination revealed that this finding appears to have low to moderate safety
significance and is being considered for escalated enforcement action in accordance with the
General Statement of Policy and Procedure for NRC Enforcement Actions (Enforcement
Policy), NUREG-1600. The current Enforcement Policy is on NRCs website at
http://www.nrc.gov/what-we-do/regulatory/enforcement/enforce-pol.html.
The plant entered a refueling outage on October 9, 2004. On October 10, the containment
equipment hatch was removed and the rail system was installed on October 11. The rail
system consisted of two sections, one installed inside containment and the other exterior to
containment. It was intended that the rail system be rapidly removed in the event of a need to
close the equipment hatch. The containment equipment hatch closure plan called for the
removal of the exterior rail only, whereas, the interior rail system was intended to be left in place
inside containment. On October 14, in preparation for lifting the reactor vessel head, an
attempt to close the equipment hatch revealed that the interior rail system interfered with and
prevented the hatch from being closed. To eliminate the interference, the interior rail system
was cut and the hatch was subsequently closed. Resolution of the issue took approximately
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
As discussed in detail in the enclosure, the significance of the finding was assessed using the
NRC Significance Determination Process (SDP). The preliminary safety significance of the
inspection finding based on the change in large early release frequency (LERF) considerations
is White.
C. Lambert -2-
Be advised that this significance assessment is preliminary. The final significance assessment
will include consideration of any further information or perspectives you provide that may
warrant reconsideration of the methodology or assumptions used during the preliminary
significance assessment.
The finding is also an apparent violation of 10 CFR 50, Appendix B, Criterion V, for failure to
have procedures in place to effect rapid removal of the interior portion of the rail system to
eliminate the interference, and is being considered for escalated enforcement action in
accordance with the General Statement of Policy and Procedure for NRC Enforcement
Actions (Enforcement Policy), NUREG-1600.
This finding did not present an immediate safety concern because core cooling was available
for the entire duration of the event and the interior rail system was modified upon discovery of
the interference so that it no longer presented an obstacle to containment equipment hatch
closure. The rail system was completely removed from the facility upon completion of the
reactor vessel head replacement activities.
Before the NRC finalizes this significance determination, 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 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.
Please contact Tom Kozak at 630-829-9866 within 10 business days of the date of 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
characterization of the apparent violation described in this letter may change as a result of
further NRC review.
C. Lambert -3-
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter
and its enclosure will be available electronically for public inspection in the NRC Public
Document Room or from the Publicly Available Records (PARS) component of NRC's
document system (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Mark A. Satorius, Director
Division Of Reactor Projects
Docket No. 50-305
License No. DPR-43
Enclosure: Significance Determination Process and
Enforcement Review Panel Background Information
cc w/encl: J. Cowan, Executive Vice President,
Chief Nuclear Officer
Plant Manager
Manager, Regulatory Affairs
J. Rogoff, Vice President, Counsel & Secretary
D. Molzahn, Nuclear Asset Manager,
Wisconsin Public Service Corporation
L. Weyers, Chairman, President and CEO,
Wisconsin Public Service Corporation
D. Zellner, Chairman, Town of Carlton
J. Kitsembel, Public Service Commission of Wisconsin
DOCUMENT NAME: G:\kewa\PSD letter - hatch issue.wpd *See previous concurrence
To receive a copy of this document, indicate in the box "C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy
OFFICE RIII RIII RIII RIII
NAME TKozak:co* KOBrien LKozak MSatorius
DATE 02/15/05 02/16/05 02/16/05 02/18/05
OFFICIAL RECORD COPY
C. Lambert -4-
ADAMS Distribution:
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ROPreports@nrc.gov (inspection reports, final SDP letters, any letter with an IR number)
SIGNIFICANCE DETERMINATION PROCESS (SDP) AND
ENFORCEMENT REVIEW PANEL
BACKGROUND INFORMATION
The significance of the finding was assessed using the SDP. Based on the initial results of the
NRC Phase 2 SDP assessment, the finding was determined to be of low to moderate safety
significance (WHITE). The applicable SDP for evaluation of this issue is the Containment SDP
which is Appendix H of Inspection Manual Chapter 0609. The finding was determined to affect
the Large Early Release Frequency (LERF) only and not Core Damage Frequency (CDF).
Therefore, it was considered to be a Type B finding. A performance deficiency existed
because an intact containment, meaning containment closure prior to reactor coolant system
(RCS) boiling, could not be maintained due to the interference of the rail system.
During the time that the interference existed, the reactor was in cold shutdown with a high
decay heat load and with the RCS vented to the containment atmosphere. This condition
represented plant operating state (POS) 2E as designated in the SDP. All emergency core
cooling systems and charging systems were available. However, during a portion of the time
that the interior rail system was installed, only one of two emergency diesel generators (EDGs)
were available. The availability of the plants mitigation capability (i.e., emergency core cooling
system (ECCS) availability, EDG availability) was determined to most closely resemble an in-
depth capability. Based on POS 2E and the assessment of an in-depth mitigation capability,
the risk significance was determined to be WHITE using the SDP.
A Phase 3 assessment was performed to determine if the Phase 2 result was bounding given
that only one EDG was available during the time the interference existed. Only the loss of
offsite power initiating event and subsequent station blackout sequence was considered. The
plant has a Technical Support Center (TSC) EDG that can provide power to a charging pump to
make up reactor vessel inventory in the event of a station blackout. This additional plant-
specific equipment was credited in the analysis. The core damage sequence of interest is a
loss of offsite power, failure of the available EDG, failure to establish charging powered by the
TSC EDG, and the failure to recover offsite power prior to core uncovery. A large early release
is assumed to occur at core uncovery because the RCS is open to containment atmosphere
and with the containment equipment hatch unable to be closed, the release is not contained.
Loss of offsite power initiating event frequencies and offsite power nonrecovery probabilities
were obtained from NUREG/CR-5496, Evaluation of Loss of Offsite Power Events at Nuclear
Power Plants: 1980 - 1996." The EDG failure probability and the TSC/charging failure
probabilities were obtained from the Kewaunee plant-specific probabilistic risk assessment
(PRA). The time to core uncovery given a station blackout was assumed to be 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
Sensitivity analyses were performed using draft NRC information on loss of offsite power events
that are soon to be published as an update to NUREG/CR-5496. Also, sensitivity analyses
were performed assuming a time to core uncovery of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The duration of the condition
was 67 hours7.75463e-4 days <br />0.0186 hours <br />1.107804e-4 weeks <br />2.54935e-5 months <br />.
Enclosure
-2-
The results of the calculations were a range of core damage frequency due to station blackout
of 4.5E-7 to 9.3E-7. The LERF was assumed to be equivalent to the CDF because the rail
system interference would prevent the closure of the equipment hatch prior to the RCS boiling.
Once boiling begins in the RCS, the analysis assumed the probability of the failure to close the
equipment hatch was 1.0. Therefore the estimated LERF also ranged from 4.5E-7 to 9.3E-7.
This estimate was compared to a base case LERF which represents no performance
deficiency. For the base case, a containment closure failure probability of .25 was assumed.
The change in LERF due to the performance deficiency then was estimated to be in the range
of 3.3E-7 to 7.0E-7.
The estimate of the probability of failure to close the equipment hatch of 1.0 was made because
the interference of the rail system was not known to plant personnel. Since the interference
was not known, no procedures existed to remove it and the method of removal was not obvious
or simple. Any actions that would be necessary to close the equipment hatch would be
severely complicated by a lack of electrical power and plant emergency conditions.
Additionally, such actions would be required under the extremely stressful situation of a station
blackout.
The licensee provided an analysis of the issue, which was similar in many respects to the
Phase 3 analysis performed by the NRC. The major difference was the assumption regarding
the likelihood of closing the equipment hatch. The licensee estimated the failure of closing the
hatch to be on the order of 6.0E-2, while the NRC estimated this probability to be 1.0 for the
reasons stated above.
In conclusion, the preliminary safety significance of the inspection finding based on the change
in LERF considerations was determined to be of low to moderate safety significance (WHITE).
If additional information is provided regarding the likelihood of closing the containment
equipment hatch or any other aspect of this analysis, it should be clearly articulated and
effectively supported.
Enclosure