IR 05000305/2005005
| ML060200325 | |
| Person / Time | |
|---|---|
| Site: | Kewaunee  |
| Issue date: | 01/20/2006 |
| From: | Louden P NRC/RGN-III/DRP/RPB5 |
| To: | Christian D Dominion Energy Kewaunee |
| References | |
| IR-05-005 | |
| Download: ML060200325 (25) | |
Text
January 20, 2006 Mr. David Senior Vice President and
Chief Nuclear Officer
Innsbrook Technical Center
5000 Dominion Boulevard
Glen Allen, VA 23060-6711SUBJECT:KEWAUNEE POWER STATION - NRC PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT NO. 05000305/2005005
Dear Mr. Christian:
On December 16, 2005, the U.S. Nuclear Regulatory Commission (NRC) completed a baseline team inspection at your Kewaunee Power Station. The enclosed report documents the
inspection findings, which were discussed on December 16 with Mr. Michael Gaffney and other
members of your staff.
The inspection was an examination of activities conducted under your license as they relate to the identification and resolution of problems, and compliance with the Commission's rules and
regulations and the conditions of your operating license. Within these areas, the inspection
involved examination of selected procedures and representative records, observations of
activities, and interviews with personnel.
Based on the samples selected for review, the inspectors identified three findings of very low safety significance (Green), two of which were determined to be violations of NRC
requirements. However, because the violations were of very low safety significance and
because the issues were entered into your corrective action program, the NRC is treating these
violations as Non-Cited Violations (NCVs), cons istent with Section VI.A of the NRC Enforcement Policy. If you contest the subject or severity of an NCV in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the
U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC
20555-0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission -
Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of
Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the
Resident Inspector Office at the Kewaunee Power Station facility.D. Christian-2-In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and any response you provide will be av ailable electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records System (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/Patrick L. Louden, Chief Projects Branch 5
Division of Reactor Projects Docket No. 50-305 License No. DPR-43
Enclosure:
Inspection Report 05000305/2005005 w/Attachment: Supplemental Information
REGION IIIDocket No:50-305 License No:DPR-43 Report No:05000305/2005005 Licensee:Dominion Energy Kewaunee, Inc.
Facility:Kewaunee Power Station Location:N490 Highway 42 Kewaunee, WI 54216Dates:November 28 through December 16, 2005 Inspectors:M. Kunowski, Project Engineer (Team Leader)
P. Higgins, Resident Inspector, Kewaunee
J. Jandovitz, Reactor Engineer
J. Neurauter, Reactor EngineerApproved by:P. Louden, Chief Projects Branch 5
Division of Reactor Projects Enclosure 1
SUMMARY OF FINDINGS
IR 05000305/2005005; Dominion Energy Kewaunee, Inc.; on 11/28/2005 - 12/16/2005;
Kewaunee Power Station; biennial baseline inspection of the identification and resolution of problems. Two violations were identified in the area of corrective actions.
The inspection was conducted by a regional projects inspector, a resident inspector, and two regional engineering inspectors. Three findings of very low safety significance (Green) were identified during this inspection, two of which were classified as Non-Cited Violations. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using
Inspection Manual Chapter (IMC) 0609, "Significance Determination Process," (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The 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.
Identification and Resolution of Problems The inspectors concluded that the licensee's implementation of its program for identifying, evaluating, and correcting nuclear safety problems was adequate. While the licensee was identifying plant problems at an appropriately low level, the inspectors had observations and one finding that indicated additional attention by plant management was warranted, particularly with the trending of conditions adverse to quality during outages to identify potentially more significant conditions or to effectively correct low-level repetitive conditions. One finding of very low safety significance was identified in this area.
In the area of prioritization and evaluation of issues, program implementation was effective, particularly with the licensee's evaluation of and corrective actions for recurrent problems with certain bistables and for operating experience related to replacement reactor head activities.
In the area of effectiveness of corrective actions, the inspectors identified two findings of very low safety significance, with associated Non-Cited Violations, for the licensee's failure to correct a procedure non-adherence issue identified during its 2004 self-assessment of the corrective action program and to correct leakage from a residual heat removal pump that could significantly increase control room and offsite doses during certain accidents. Leakage from the residual heat removal pumps has been an issue at Kewaunee since 1979.
From interviews conducted during this inspection and a review of corrective action program and employee concerns program documents, the in spectors concluded that workers at Kewaunee felt free to input nuclear safety findings into the corrective action program or the employee concerns program. A.Inspector-Identified and Self-Revealed Findings
Cornerstone: Mitigating Systems
- Green.
The inspectors identified a finding of very low safety significance for the licensee not reviewing corrective action program documents (CAPs) during outages for potential trends of conditions adverse to quality. As part of the screening process of CAPs, the licensee assigned, as possible, CAPs to various "hot buttons." Hot buttons were searchable categories in the corrective ac tion program computer system that had been established for various problems, such as equipment tagging errors, security door control, and reactivity management. For non-outage times, the licensee assigned a monthly number of hits for each hot button that, if exceeded for 3 months in succession, would result in the generation of a CAP to investigate a possible trend. However, as of
December 16, 2005, the licensee did not use hot button action levels during outages when the number of CAPs written was much higher than during non-outage times.
This finding is greater than minor because if left uncorrected would become a more significant safety concern. This finding is not suitable for Significance Determination
Process evaluation, but has been reviewed by NRC management and is determined to be a finding of very low safety significance. The cause of the finding is related to the cross-cutting element of problem identification and resolution, because of not identifying potential conditions adverse to quality through trending of CAPs during outages.
(Section 4OA2a.(2)(i))*Green. The inspectors identified a finding of very low safety significance and a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," for the failure to take corrective action for procedure non-compliance identified during the licensee's 2004 self-assessment of the corrective action program. As a result of the assessment, CAP025194, "Corrective Action Program Procedure and Guidance
Document Use," was written and documented that plant workers were not following corrective action program procedures for apparent cause evaluations and root cause evaluations, effectiveness review content, priority and due date assignments, initiator feedback, and documentation of corrective action completion. To correct this problem, corrective action CA018094, "Corrective Action Program Procedure and Guidance
Document Use," was written and specified one or 2 weeks of requiring "in-hand" use by the plant staff of the corrective action program administrative procedure. However, completion of this action was delayed several times and on July 25, 2005, CAP025194 and CA018094 were closed with the only documented action taken being a
July 18, 2005, meeting of the station human performance steering committee at which the licensee decided not to take action because of the pending transition to the corrective action program documents of the plant's new owner.
This finding is greater than minor because if left uncorrected would become a more significant safety concern. This finding is not suitable for Significance Determination
Process evaluation, but has been reviewed by NRC management and is determined to be a finding of very low safety significance. The cause of the finding is related to the cross-cutting element of problem identification and resolution, because of the failure to take corrective action for non-adherence to station procedures. (Section 4OA2c.(2)(i))*Green. A finding of very low safety significance that was a Non-Cited Violation of10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified for the licensee's ineffective corrective action to repair a leak on the seal of the "B" residual heat removal (RHR) pump. The leak was identified on November 2, 2005, when the pump was stopped following the performance of a required surveillance. The leak rate exceeded leakage control program limits. A similar leak was identified on
June 16, 2004, for which the licensee replaced the seal in November 2004.
3 This finding is greater than minor because it was associated with the "RCS (reactor coolant system) equipment and barrier performance" attribute of the barrier integrity cornerstone and does affect the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Although the RCS barrier was affected, the finding did not affect the mitigation capability of the RHR system and did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator or affect the containment integrity.
Therefore, the finding is of very low safety significance. (Section 4OA2c.(2)(ii))
B.Licensee-Identified Violations
None.
4
REPORT DETAILS
OTHER ACTIVITIES (OA)
4OA2Problem Identification and Resolution a.Effectiveness of Problem Identification (1)Inspection Scope The inspectors reviewed items selected from the cornerstones of safety to determine if problems were being properly identified, characterized, and entered into the corrective
action program for evaluation and resolution. At Kewaunee, problems entered into the
program are documented as CAPs (i.e., condition reports). Included in this review were
numerous CAPs, the reports of the 2004 and 2005 Kewaunee self-assessments, Nuclear Oversight (quality assurance) reports, Licensee Event Reports (LERs), other
plant documents, and previous NRC inspection reports. The inspectors also conducted
plant tours and interviewed plant personnel to identify equipment or process problems
that had not been entered into the corrective action program. The previous NRC
problem identification and resolution team inspection was conducted at the end of 2003 (Inspection Report (IR) 05000305/2003010).
(2)Assessment The inspectors concluded that the licensee was generally effective in identifying problems and the threshold of the majority of plant personnel was appropriately low.
Both the previous operating company, t he Nuclear Management Company, LLC (NMC)and the current owner and operator (as of July 5, 2005), Dominion, emphasized to plant
personnel a low threshold for documenting problems in CAPs. The number of CAPs
generated in recent years have indicated that this emphasis has been effective:*in 2001, 4903 CAPs (outage year, including steam generator replacement),*in 2002, 3867 CAPs (non-outage year),
- in 2003, 5208 CAPs (outage year),
- in 2004, 5367 CAPs (outage year), and
- in 2005, 5773 CAPs (outage year).
The inspectors, however, had several observations, including one finding, that indicated additional effort was warranted in the area of identification of problems. These
observations are discussed below.(i)Trending During Outages Introduction
- The inspectors identified a finding of very low safety significance (Green) for the licensee not reviewing CAPs during outages for potential trends
of conditions adverse to quality.
5 Description
- As part of the screening process of CAPs, the licensee assigns, as possible, CAPs to various "hot buttons." Hot buttons are searchable categories
in the corrective action program comput er system that have been established for various problems, such as equipment tagging errors, security door control, and
reactivity management. For non-outage times, the licensee assigned a monthly
number of hits for each hot button that, if exceeded for 3 months in succession, would result in the generation of a CAP to investigate a possible trend. However, the licensee does not use hot button action levels during outages. The
explanation was that outages were known to result in more CAPs being
generated and that developing appropriate action levels for the known increase
in CAPs in the various hot button categories would be problematic. The
inspectors concluded that timely hot button categorization and analysis during
outages could help prevent a significant program, process, or work group
problem that was currently showing up as lower level issues or could reduce or
eliminate repeat lower level issues.
Analysis:
The inspectors determined that the licensee's failure to review CAPs during outages to identify and address potential trends in conditions adverse to
quality was a licensee performance deficiency warranting a significance
evaluation. The inspectors concluded that the finding was more than minor in
accordance with Inspection Manual Chapter (IMC) 0612, "Power Reactor
Inspection Reports," Appendix B, "Issue Screening," issued September 30, 2005, in that, the finding if left uncorrected would become a more significant safety
concern. This finding (FIN 05000305/2005005-01) is not suitable for Significance
Determination Process (SDP) evaluation, but has been reviewed by NRC
management and is determined to be of very low safety significance (Green).
The inspectors also determined that the finding affected the cross-cutting element of problem identification and resolution, because of not identifying
potential conditions adverse to quality through trending of CAPs during outages.
Enforcement
- No violation of NRC requirements was identified for this finding.
The licensee entered the issue into the corrective action program as
CAP030559.(ii)Additional Observations on Trending and Identification of Problems*The licensee's 2004 self-assessment of the corrective action program identified that the trend analysis program was adding little value; however, this was not
documented in a CAP. The explanation gi ven in the assessment report implied that one was not needed because the recently developed monthly department
"roll-up" process (DRUM process) would likely address the problems after
several months of run time. Among other items, the DRUM process had the
various departments assess department-related CAPs from the past month to
identify any trends. The inspectors concluded that a CAP should have been
written to ensure that the DRUM process was reviewed after several months to
assess if the original issue from the 2004 assessment was addressed. With the
imminent change from NMC to Dominion and a forced outage from February 20
to July 2, 2005, the DRUM process was not implemented. A similar Dominion 6 process has since been implemented.*Trend program concerns were again identified as part of the 2005 self-assessment of the corrective action program. The concerns, identified by an
NOS evaluator (Nuclear Oversight-quality assurance) were documented in
CAP029587, "The CAP Trending Program Expectations Not Met." Issues
documented included not using hot button trends during non-refueling outages;
the last published quarterly corrective action program trend report was for the
third quarter of 2004, as of the third quarter of 2005; and a backlog of 343 existed
for a final corrective action program quality check of completed apparent cause
evaluations and conditions evaluations. *At a CAP screening meeting attended by the inspectors, the screening team did not question why CAP030351 had just been written, on November 30, 2005, for
an issue regarding the use of current procedures in training that was first
identified on March 29. The licensee wrote CAP030543 to follow-up on the
inspectors' observation. *The resident inspectors, in their daily review of newly written CAPs, have identified instances where more than one potential issue was documented in a
CAP. The licensee's policy is "one issue, one CAP." The licensee entered this
apparent discrepancy into its corrective action program as CAP030560 for
evaluation.
b.Prioritization and Evaluation of Issues (1)Inspection Scope The inspectors reviewed the licensee's significance classification and evaluation of a sample of CAPs, apparent cause evaluations (ACEs), and root cause evaluations (RCEs). The inspectors' assessment included a review of the following attributes:
significance category assigned to a CAP, the adequacy of operability and reportability
determinations, the extent of condition evaluations, and the appropriateness of using
whatever causal investigation was used. The licensee's prioritization and evaluation of
selected operating experience issues regarding reactor vessel head lifting, in
Westinghouse technical bulletins, and with Foxboro instruments were also assessed by
the inspectors.
The inspectors also attended several CAP daily screening meetings and a corrective action review board meeting where ACEs and RCEs were reviewed by licensee
management. At these meetings, the inspectors assessed the licensee's evaluation of
issues in CAPs, ACEs, and RCEs. (2)Assessment(i)ACEs and RCEs
- For the ACEs and RCEs reviewed, no significant problems were identified by the inspectors. The causes identified by the licensee
appeared appropriate and the identified corrective actions, if fully implemented, should correct the problems that caused the original issue.
7(ii)Operating Experience
- For CAPS associated with external operating experience contained in Westinghouse technical bulletins, the inspectors did not identify any
significant problems but did have some observations regarding thoroughness of
the documentation of evaluation results. The licensee indicated that these
observations would be evaluated and appropriate corrective actions would be
taken, as necessary. *For Operating Experience OE002555, "CROSSFLOW Ultrasonic Flow Measurement System Performance Ob servations," three action items had been assigned to change three procedures. One procedure change had
been completed, one procedure deleted with no explanation in the file as
to whether another procedure had taken its place; and one change had
not yet been made. In addition, the OE evaluation mentioned the need
for additional training but no action item was assigned to provide the
training.*For OE002902, "RCP Motor Recommended 1-Year, 5-Year and 10-Year Maintenance," no basis was given for recommendations that would not be
followed. *For OE005168, "Updated Reactivity Surveillance Policy for B-10 Isotopic Concentration," the licensee stated that the recommended actions would
be taken, but the frequency of the surveillance was different from that
contained in the Bulletin and no basis for the difference was provided.(iii)Foxboro Instruments:
The inspectors reviewed the licensee's evaluation of and subsequent corrective actions for failures of certain Foxboro bistables. The
licensee's efforts were in followup to observations made by the resident
inspectors (IR 05000305/2005008). The resident inspectors had determined that
the licensee did not always address potential operability considerations when
bistables associated with safety-related Technical Specification systems were
found out-of-tolerance in the non-conservative direction. Also, the resident
inspectors identified that because the licensee considered Foxboro instruments
in their own Maintenance Rule system, out-of-tolerance bistables were handled
through the Maintenance Rule process and components that had out-of-
tolerance bistables were not individually evaluated. Lastly, the inspectors had
determined that when an out-of-tolerance condition was identified, a thorough
extent of condition was not always performed.
During the current inspection, the inspectors reviewed the history of the Foxboro bistable failures contained in the corrective action program, reviewed the results
and the trending data of the surveillance procedure of the safety-related Foxboro
instruments for approximately the last 1 1/2 years, and discussed these results
with station personnel. The inspectors determined through discussions with the
Maintenance Rule engineer that Foxboro instrument failures were now evaluated
against the plant system affected by the Foxboro instrument as well as the
simulated Foxboro system. Unavailab ility, if any, caused by the Foxboro instrument failure was being logged against the plant system. The results of the
monthly surveillance procedures for the last 12-18 months for the safety-related 8 bistables were reviewed with the instrument and control (I&C) engineer. In most cases, the failure of the bistable could be predicted with the trending data. The
Foxboro instrument would be recalibrated or replaced prior to the predicted
failure. In some cases, the data were not predictable due to large variances, but
still within acceptance criteria. The I&C engineer was now engaged with all the
Foxboro surveillances conducted and the results and operability determinations
resulting from failures of safety-related instruments. In addition, the I&C engineer
provided information on the replacement program for the Foxboro instruments which should be completed in 2006. The inspectors determined from this
information that the corrective actions taken since the previous inspection
adequately addressed the previous concerns of the resident inspectors.(iv)Reactor Head Drop Analysis
- During the fall 2004 refueling outage, Kewaunee installed a new reactor vessel head that weighed less than the original reactor
vessel head. The effect of replacement reactor vessel head weight on the
original head drop analysis was evaluated by the licensee in its 10 CFR 50.59
analysis. The inspectors considered the original head drop analysis to be
bounding and conservative for the lower weight, replacement reactor vessel
head. Therefore, a review of the original head drop analysis was not performed
by the inspectors.
In later inspections at other licensees where replacement heads weighed more than the original heads, non-conservative assumptions and methodologies and
incomplete resolution of load drop analysis results were identified for head drop
analyses, as described in NRC Regulatory Issue Summary (RIS) 2005-25, "Clarification of NRC Guidelines for Control of Heavy Loads," dated
October 31, 2005. In addition, RIS 2005-25 also clarified NRC regulatory
guidelines for the control of heavy loads to assure the safe handling of heavy
loads in areas where a load drop could impact stored spent fuel, fuel in the
reactor core, or equipment that may be required to achieve safe shutdown or
permit continued decay heat removal.
During the current inspection, the inspectors reviewed the licensee's evaluation and corrective actions pertaining to industry operating experience and
RIS 2005-25 related to its reactor vessel head drop analysis and control of heavy
loads. The evaluation of operating experience had been entered into the
licensee's corrective action program as CAP027482. The licensee's review
identified that its reactor vessel head drop analysis used the same non-
conservative method of analysis as the Prairie Island Nuclear Generating Plant (IR 05000282/2005004; 05000306/2005004 (ML052020420) dated
July 21, 2005). The inspectors verified that the licensee's corrective action, CA019697, included a plan to update the head drop analysis using finite element
methods based on a "conservation of energy" methodology. The updated head
drop analysis will use a heavier weight, consistent with a head assembly upgrade
package. CA019697 indicated that the licensee's goal was to update the head
drop analysis prior to the fall 2006 refueling outage. Licensee senior
management confirmed this in a discussion with the inspectors.
The inspectors interviewed knowledgeable licensee staff to determine the 9 potential safety significance of the non-conservative methodology used in its current head drop analysis. The licensee's staff indicated that the Kewaunee
reactor vessel support design was very similar to that of Prairie Island, and the
results from the revised Prairie Island head drop analysis using finite element
methods gave reasonable assurance that the current lighter weight Kewaunee
head (approximately 140,000 pounds versus 200,000 pounds for Prairie Island)
could be safely lifted above the reactor vessel to an elevation necessary to
remove and replace the head during refueling operations.
The inspectors observed that current licensee procedures pertaining to removal and replacement of the reactor vessel head did not contain a maximum head lift
height restriction. The inspectors noted that licensee procedures may need to be
revised to specify a maximum head lift height restriction to be consistent with
results from the updated head drop analysis.
The inspectors concluded that industry operating experience and NRC issues identified in RIS 2005-25 related to Kewaunee's reactor vessel head drop
analysis and control of heavy loads have been identified by the licensee, entered
into its corrective action program, and corrective actions specified and scheduled
to resolve concerns and issues related to the current head drop analysis prior to
the fall 2006 refueling outage.
c.Effectiveness of Corrective Actions(1)Inspection Scope The inspectors reviewed selected CAPs and associated corrective actions (CAs)to evaluate the effectiveness of the licensee's corrective actions taken for issues.
The inspectors reviewed condition evaluations (CEs), ACEs, and RCEs to
determine if corrective actions, commensurate with the significance of the issues, were identified and implemented in a timely manner, including corrective actions
to address longstanding or repetitive issues.
The inspectors also verified the continued implementation of a sample of completed corrective actions. The sample that was selected for review was
based, in part, on the safety and risk significance of the issues pertaining to the
reactor safety strategic performance area. Included in the review by the
inspectors were corrective actions taken for licensee self-assessment findings, issues in licensee event reports (LERs), and for Non-Cited Violations (NCVs)
discussed in previous NRC inspection reports. (2)Assessment For most of the issues reviewed by the inspectors, appropriate and timely corrective actions were taken; however, as discussed below, two findings of very
low safety significance involving violations of NRC requirements were identified
by the inspectors.(i)Corrective Action Not Taken 10 Introduction
- The inspectors identified a finding of very low safety significance (Green) for the failure to take corrective action for an
issue regarding procedure compliance identified during the licensee's
2004 self-assessment of the corrective action program.
Description
- In the 2004 self-assessment of the corrective action program, one of the four CAPs written for identified problems was
CAP025194, "Corrective Action Program Procedure and Guidance
Document Use," January 27, 2005. This CAP documented that plant
workers were not following corrective action program procedures and
guidance documents (essentially, NMC procedures and documents)
for ACEs and RCEs, effectiveness review content, priority and due
date assignments, initiator feedback, and documentation of corrective
action completion. To correct this problem, CA018094, "Corrective
Action Program Procedure and Guidance Document Use," was written
and specified 1 or 2 weeks of requiring "in-hand" use by the plant staff
of the corrective action program administrative procedure (General Nuclear Procedure GNP-11.08.01, "Action Request Process") in
February-March 2005. However, completion of this action was
delayed several times and on July 25, 2005, CAP025194 and
CA018094 were closed with the only documented action taken being
a July 18, 2005, meeting of the station human performance steering
committee. At this meeting, it was decided that the "in-hand"
procedure use recommendation would not be implemented because
training would be provided to plant staff on standards and
expectations of procedure use and adherence when the Dominion
fleet corrective action program was implemented at Kewaunee.
During the current inspection, licensee representatives stated that the
Dominion corrective action program procedure was expected to be
implemented in late December 2005 or January 2006.
Although no specific corrective action was taken for this self-assessment problem, the licensee had emphasized corrective action
program procedure adherence to the plant staff in periodic plant
newsletters and daily alignment meetings ("D-15 meetings"). The
inspectors noted that several of the seven issues identified during the
2005 self-assessment of the corrective action program were caused, in part, by plant staff not following correction program procedures and
guidance documents.
Analysis:
The inspectors determined that the licensee's failure to take corrective action to address plant staff failure to follow the corrective
action program administrative procedure was a licensee performance
deficiency warranting a significance evaluation. The inspectors
concluded that the finding was more than minor in accordance with
IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue
Screening," issued September 30, 2005, in that, the finding if left
uncorrected would become a more significant safety concern. This
finding is not suitable for SDP evaluation, but has been reviewed by 11 NRC management and is determined to be a finding of very low safety significance (Green).
The inspectors also determined that the finding affected the cross-cutting element of problem identification and resolution (corrective
action), because of the failure to take corrective action for non-
adherence to station procedures.
Enforcement:
10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," requires, in part, that measures be established to assure that
conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, and nonconformances are promptly identified
and corrected. Contrary to this, as of December 16, 2005, the
licensee had not corrected a condition adverse to quality, the failure
by plant staff to follow corrective action program procedures that was
identified during the 2004 self-assessment of the corrective action
program. Because this finding was of very low safety significance (Green) and because it had been entered into the corrective action
program (as CAP030538), it is being treated as an NCV, consistent
with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000305/2005005-02). (ii)Inadequate Corrective Action Taken Introduction
- The inspectors identified a finding of very low safety significance (Green) for the failure to take adequate corrective action
for a leaky seal on a residual heat removal (RHR) pump. An NCV had
previously been identified for the leaky seal during a mid-2004 NRC
inspection.
Description
- On June 16, 2004, during a routine quarterly surveillance, the licensee identified that the seal of the 'B' RHR
pump was leaking excessively after the pump was stopped. The
licensee estimated the leakage was approximately 1 gallon per minute (gpm) or 60 gallons per hour (gph). This was greater than the 6-gph
emergency core cooling system leakage allowed by the
System Integrity Program (SIP), as referenced by Technical
Specification 6.12, and greater than the 12-gph leakage discussed in
Chapter 14 of the Updated Safety Analysis Report (USAR) for
calculation of control room and offsite doses. The licensee entered a
7-day administrative Limiting Condition for Operation per the SIP and
the licensee declared the pump operable but degraded, on the basis
that the mechanical seal stopped leaking after the pump was
electrically started and stopped in short succession (i.e., "bumped").
The NRC resident inspectors determined that excessive seal leakage
had occurred on numerous occasions in the past and previous actions
had not been effective in correcting this condition adverse to quality.
An NCV (05000305/2004004-01) for failure to correct a condition
adverse to quality was identified and was documented in IR 2004004, 12 dated July 29, 2004. The licensee had documented the problem in its corrective action program as CAP021589 and CAP021744. For its
corrective action, the licensee replaced the seal in November 2004
during a refueling outage.
During the current inspection, the inspectors reviewed the effectiveness of the corrective action for the 2004 leak and identified
that on November 2, 2005, the "B" RHR pump replacement seal
leaked when the pump was stopped during a routine quarterly
surveillance. As in June 2004, operators stopped the leak by
"bumping" the pump. For the subsequent operability evaluation, the
licensee estimated that the leakage was less than 1 gpm-the leakage
had not been measured before the pump was bumped. The shift
manager declared the pump operable, on the basis that the leakage
stopped when the pump was "bumped" and that the radiological
analysis for the June 2004 leak, which assumed a 60-gph leak rate, determined that there was no significant impact on control room or
offsite doses.
In response to questions by the NRC inspectors, the licensee re-estimated the leakage on November 2, 2005, as greater than 6 gph
but less than 60 gph, a rate in excess of that allowed by the SIP. The
inspectors also noted that the initial operability evaluation for the leak
in June 2004 did not address the potential radiological consequences
of the RHR system barrier leaking reactor coolant outside
containment in excess of SIP and USAR limits. For the operability
evaluation for the November 2005 leak, the licensee reviewed the
potential impact of the estimated leakage on control room and offsite
doses and demonstrated that no dose limits were likely exceeded.
From interviews and a review of corrective action program records and work orders, the inspectors determined that leakage from the
RHR pump seals on both trains had occurred numerous times since
1979 following the shutdown of the pumps. Historically, the licensee
stopped the leakage by rotating the pump shaft, either electrically or
manually, until the leak stopped. This method had been incorporated
in Procedure A-MDS-30, "Miscellaneous Drains and Sumps (MDS)
Abnormal Operation," November 22, 2005. Section 4.10, "RHR Pump
Pit Sump," Step 2.a., stated, "IF RHR pump was NOT running, THEN seal leakage may be stopped by rotating shaft by hand or bumping
motor."
Analysis:
The inspectors determined that the licensee's failure to take effective corrective actions to address the RHR pump seal leakage
was a performance deficiency warranting a significance evaluation.
This self-revealed finding was greater than minor because the finding
was associated with the "RCS (reactor coolant system) equipment
and barrier performance" attribute of the barrier integrity cornerstone
and does affect the cornerstone objective of providing reasonable 13 assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide
releases caused by accidents or events.
The inspectors determined that the finding could not be evaluated using the SDP. Although the inspectors, with the assistance of a
Region III Senior Reactor Analyst, determined that the RCS barrier
was affected, the Phase 2 worksheets were not applicable because
this issue did not affect the mitigation capability of the RHR system.
The finding also did not contribute to the likelihood of a primary or
secondary system loss of coolant accident initiator or affect the
containment integrity. Therefore, this finding was reviewed by a
Region III Branch Chief in accordance with IMC 0612, Section 05.04c, who agreed with the inspectors that this finding was of very low safety
significance (Green).
The inspectors also determined that the finding affected the cross-cutting element of problem identification and resolution, because of
the failure to take effective corrective action to address the RHR pump
seal leakage.
Enforcement:
Title 10 CFR Part 50, Appendix B, Criterion XVI,"Corrective Action," requires that measures 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. Contrary to
this requirement, as of December 16, 2005, actions taken to correct a
leaky seal on the "B" RHR pump, a condition adverse to quality, have
not been effective. Because this finding was of very low safety
significance (Green) and because it had been entered into the
corrective action program (as CAP030527, on December 14, 2005), it
is being treated as an NCV, consistent with Section VI.A.1 of the NRC
Enforcement Policy (NCV 05000305/2005005-03).
During the review of this issue, the inspectors questioned 1) the basis for a 2 gph limit on RHR train leakage that previously was in the
Technical Specifications and was a basis for the current 6 gph limit in
the SIP and, 2) whether the licensee had properly transferred all
requirements to the SIP and other administratively controlled
documents when the NRC approved (on February 25, 1998)
Kewaunee's implementation of Option B of Appendix J, "Primary
Reactor Containment Leakage Testing for Water-Cooled Power
Reactors," of 10 CFR 50. The licensee could not answer the
questions during the inspection and, consequently, the resident
inspectors will follow-up as part of their routine inspection activities. d.Assessment of Safety-Conscious Work Environment(1)Inspection Scope 14 To determine if plant personnel were reluctant to raise nuclear safety concerns, the inspectors questioned workers in the plant and interviewed the corporate
manager (and recent site manager) of the station employee concerns program.
The inspectors also reviewed program records to determine if employee
concerns had been properly evaluated and corrected, as necessary.(2)Assessment The inspectors concluded that licensee personnel were willing to raise safety concerns and that nuclear safety issues raised to the employee concerns
program were properly evaluated and corrected. 4OA6Meetings On December 16, 2005, the team presented the preliminary inspection results to Mr. M. Gaffney and other members of the licensee's staff, who acknowledged the
findings. The licensee did not identify any information, provided to or reviewed by the
team and likely to be included in the inspection report, as proprietary.4OA7Licensee-Identified Violations None.ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- L. Armstrong, Director of Engineering
- R. Bower, Technical Specialist, Corrective Actions
- T. Breene, Manager, Nuclear Licensing
- K. Davison, Director, Nuclear Station, Operations and Maintenance
- M. Gaffney, Site Vice-President
- D. Gauthier, Nuclear Quality Specialist
- K. Hoops, Site Director
- W. Hunt, Maintenance Manager
- R. Nicolai, Organizational Effectiveness Manager
- K. Peckham, Nuclear Oversight Manager
- K. Peveler, Manager Engineering Programs
- D. Sieracki, Dominion Fleet Manager, Employee Concerns Program
- T. Taylor, Licensing and Compliance Group
- T. Van Valkenburg, Technical Specialist, Corrective Actions
- T. Webb, Director of Safety and Licensing
Nuclear Regulatory Commission
- S. Burton, Senior Resident Inspector, Kewaunee
- P. Louden, Chief, Reactor Projects Branch 5
- M. Satorius, Director, Division of Reactor Projects
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
05000305/2005005-01FINNo Trending of Adverse Conditions Identified During
Outages (Section 4OA2a.(2)(i))05000305/2005005-02NCVFailure to Correct Procedure Non-Adherence (Section 4OA2c.(2)(i))05000305/2005005-03NCVFailure to Adequately Correct Residual Heat Removal
Pump Seal Leakage (Section 4OA2c.(2)(ii))
Closed
05000305/2005005-01FINNo Trending of Adverse Conditions Identified During
Outages (Section 4OA2a.(2)(i))05000305/2005005-02NCVFailure to Correct Procedure Non-Adherence (Section 4OA2c.(2)(i))05000305/2005005-03NCVFailure to Adequately Correct Residual Heat Removal
Pump Seal Leakage (Section 4OA2c.(2)(ii))
Discussed
05000305/2004004-01NCVNon-Cited Violation of 10 CFR Part 50, Appendix B,Criterion XVI, "Corrective Action," for the Failure to
Correct Historical Residual Heat Removal Pump
Mechanical Seal Leakage (Section 4OA2c.(2)(ii))