ML112550639
| ML112550639 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 09/12/2011 |
| From: | Powers D Division of Reactor Safety IV |
| To: | Flores R Luminant Generation Co |
| References | |
| IR-11-006 | |
| Download: ML112550639 (36) | |
See also: IR 05000445/2011006
Text
UNITED STATES
NU CLEAR REG ULAT O RY CO M M I SSI O N
R E GI ON I V
612 EAST LAMAR BLVD, SUITE 400
ARLINGTON, TEXAS 76011-4125
September 12, 2011
Mr. Rafael Flores, Senior Vice President
and Chief Nuclear Officer
Luminant Generation Company, LLC
Comanche Peak Nuclear Power Plant
P.O. Box 1002
Glen Rose, TX 76043
SUBJECT: COMANCHE PEAK NUCLEAR POWER PLANT - NRC PROBLEM
IDENTIFICATION AND RESOLUTION INSPECTION
REPORT 05000445/2011006 AND 05000446/2011006
Dear Mr. Flores:
On July 28, 2011, the U. S. Nuclear Regulatory Commission (NRC) completed a team
inspection at Comanche Peak Nuclear Power Plant. The enclosed report documents the
inspection findings, which were discussed on July 28, 2011, with Mr. Mitch Lucas, Site Vice
President, and other members of your staff.
The inspection examined activities conducted under your license as they relate to identification
and resolution of problems, safety and compliance with the Commissions rules and regulations
and with the conditions of your operating license. The team reviewed selected procedures and
records, observed activities, and interviewed personnel. The team also interviewed a
representative sample of personnel regarding the condition of your safety conscious work
environment. The team concluded that in general, problems were properly identified, evaluated,
and corrected.
This report documents four NRC-identified findings of very low safety significance (Green). All
of these findings were determined to involve violations of NRC requirements. However,
because of the very low safety significance of the violations and because they were entered into
your corrective action program, the NRC is treating these violations as noncited violations
consistent with Section 2.3.2 of the NRC Enforcement Policy. If you contest these noncited
violations, or the significance of the noncited violations, you should provide a response within 30
days of the date of this inspection report, with the basis for your denial, to the Nuclear
Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with
copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 612 E.
Lamar Blvd., Suite 400, Arlington, Texas, 76011-4125; the Director, Office of Enforcement,
United States Nuclear Regulatory Commission, Washington DC 20555-0001; and the NRC
Resident Inspector at the Comanche Peak Nuclear Power Plant. In addition, if you disagree
with the crosscutting aspect assigned to any finding in this report, you should provide a
response within 30 days of the date of this inspection report, with the basis for your
disagreement, to the Regional Administrator, Region IV, and the NRC Resident Inspector at the
Comanche Peak Nuclear Power Plant.
Luminant Generation Company, LLC -2-
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its
enclosure, and your response (if any) 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
www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Dr. Dale A. Powers, Acting Chief
Technical Support Branch
Division of Reactor Safety
Dockets: 50-445; 50-446
Enclosure:
Inspection Report 05000445/2011006 and 05000446/2011006
w/Attachments: Attachment 1, Supplemental Information
Attachment 2, Initial Information Request
cc w/ Enclosure:
Distribution via Listserv for Comanche Peak
Luminant Generation Company, LLC -3-
Electronic distribution by RIV:
Regional Administrator (Elmo.Collins@nrc.gov)
Deputy Regional Administrator (Art.Howell@nrc.gov)
DRP Director (Kriss.Kennedy@nrc.gov)
Acting DRP Deputy Director (Jeff.Clark@nrc.gov)
DRS Director (Anton.Vegel@nrc.gov)
DRS Deputy Director (Tom.Blount@nrc.gov)
Senior Resident Inspector (John.Kramer@nrc.gov)
Branch Chief, DRP/A (Wayne.Walker@nrc.gov)
Senior Project Engineer, DRP/A (David.Proulx@nrc.gov)
Project Engineer, DRP/A (Christopher.Henderson@nrc.gov)
CP Administrative Assistant (Sue.Sanner@nrc.gov)
Public Affairs Officer (Victor.Dricks@nrc.gov)
Public Affairs Officer (Lara.Uselding@nrc.gov)
Project Manager (Balwant.Singal@nrc.gov)
Branch Chief, DRS/TSB (Dale.Powers@nrc.gov)
RITS Coordinator (Marisa.Herrera@nrc.gov)
Regional Counsel (Karla.Fuller@nrc.gov)
Congressional Affairs Officer (Jenny.Weil@nrc.gov)
OEMail Resource
ROPreports
RIV/ETA: OEDO (John.McHale@nrc.gov)
DRS/AA (Loretta.Williams@nrc.gov)
ADAMS: No x Yes SUNSI Review Complete Reviewer Initials: DAP
x Publicly Available x Non-Sensitive
Non-publicly Available Sensitive
RI:DRP/D SRI:DRP/A SRI:DRS/EB1 C:DRP/A
JPReynoso JGKramer WCSifre WCWalker
/RA/ /RA/ /RA/ /RA/
09/12/2011 09/07/2011 09/01/2011 09/07/2011
RI:DRP/C AC:DRS/TSB
JDBraisted DAPowers
/RA/ /RA/
09/01/2011 09/12/2011
OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket: 05000445, 05000446
Report: 05000445/2011006 and 05000446/2011006
Licensee: Luminant Generation Company, LLC
Facility: Comanche Peak Nuclear Power Plant
Location: FM-56, Glen Rose, Texas
Dates: July 11 through July 28, 2011
Team Leader: J. Reynoso, Resident Inspector
Inspectors: J. Kramer, Senior Resident Inspector
W. Sifre, Senior Reactor Inspector
J. Braisted, Reactor Inspector
Approved By: Dr. Dale A. Powers, Acting Chief
Technical Support Branch
Division of Reactor Safety
-1- Enclosure
SUMMARY OF FINDINGS
IR 05000445/2011006, 05000446/2011006; 07/11/2011-7/28/2011; Comanche Peak Nuclear
Power Plant "Biennial Baseline Inspection of the Identification and Resolution of Problems."
The inspection was performed by reactor inspectors and resident inspectors. Four noncited
violations of very low safety significance (Green) were identified during this inspection. The
significance of most findings is indicated by their color (Green, White, Yellow, 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 4, dated December 2006.
Identification and Resolution of Problems
The team reviewed approximately 450 condition reports, work orders, engineering evaluations,
root and apparent cause evaluations, and other supporting documentation to determine if
problems were being properly identified, characterized, and entered into the corrective action
program for evaluation and resolution. The team reviewed a sample of system health reports,
self-assessments, trending reports and metrics, and various other documents related to the
corrective action program. Based on these reviews, inspection team concluded that the
implementation of the corrective action program at Comanche Peak Nuclear Power Plant Units
1 and 2 is acceptable. The team noted that the licensee personnel were identifying issues at a
sufficiently low threshold, evaluating, prioritizing problems, and generally analyzed operating
experience appropriately. The team determined that licensee personnel were performing
effective self-assessments, and have maintained an effective safety conscious work
environment.
The team identified challenges in the area of effective corrective actions and evaluation of
problems. The team noted that the licensee has long-standing equipment problems, which may
indicate lack of effective corrective actions. The team determined that ineffective corrective
actions for diesel generator cam cover bolts, jacket water leaks, service water vacuum breakers
and globe valves (HermaValves) continued.
The team also determined the licensee staff appropriately evaluated industry operating
experience for relevance to the facility and had entered applicable items in the corrective action
program. The licensee generally used industry operating experience when performing root
cause and apparent cause evaluations. However, the team noted that sometimes these actions
were not thorough. As an example, the team determined there was adequate information from
industry operating experience, to prevent the failure of motor operated valves due to use of dry
stem lubricant. The licensee staff implemented most of the needed actions, but due to
scheduling and inaccessibly, failed to appropriately correct the condition which resulted in a
motor operated valve not performing its safety function.
-2- Enclosure
A. NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
- SL-IV. The team identified a Severity Level IV noncited violation of
10 CFR 50.59, Changes, Tests, and Experiments, associated with the failure to
conclude that a change to the Final Safety Analysis Report required prior NRC
review and approval prior to implementation. Specifically, the licensee made
changes to the Final Safety Analysis Report that resulted in more than a minimal
increase in the likelihood of occurrence of a malfunction of a structure, system, or
component important to safety. The licensee entered the finding in the corrective
action program as Condition Report CR 2011-008509.
This finding was more than minor because there was a reasonable likelihood that
the change would require a prior NRC approval. Violations of 10 CFR 50.59 are
violations that potentially impede or impact the regulatory process and are
processed through traditional enforcement. As required by Section 7.3 of the
Enforcement Policy, the team performed a Phase 1 screening in accordance with
Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and
Characterization of Findings, to determine the significance of the finding. The
team determined that the finding is of very low safety significance (Green)
because the finding: (1) was not a design or qualification issue confirmed not to
result in a loss of operability or functionality; (2) did not represent an actual loss
of safety function of the system or train; (3) did not result in the loss of one or
more trains of nontechnical specification equipment; and (4) did not screen as
potentially risk significant due to a seismic, flooding, or severe weather initiating
event. Since violations of 10 CFR 50.59 may result in conditions evaluated as
having very low safety significance by the Significance Determination Process,
the team categorized the finding as Severity Level IV in accordance with the
Enforcement Manual. The finding was a violation determined to be of very low
safety significance, was not repetitive or willful, and was entered into the
corrective action program. Therefore, this violation is being treated as a noncited
violation consistent with the NRC Enforcement Policy. The team did not identify
a crosscutting aspect with this finding since this performance issue occurred in
2004 and is not reflective of current performance (Section 4OA2.5a).
- Green. The team identified a Green noncited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the
licensees failure to have documented instructions for an activity affecting quality.
Specifically, the licensee did not have documented instructions for filling the
diesel generator jacket water system when the normal fill method would not be
available during a loss-of-offsite power. Prior to July 27, 2011, the licensee failed
to have adequate instructions for filling the diesel generator jacket water system,
an activity affecting quality, during a loss-of-offsite power. The licensee entered
the finding into the corrective action program as Condition Report
CR 2011-008510.
-3- Enclosure
This performance deficiency was determined more than minor because it was
associated with the procedure quality attribute of the Mitigating Systems
Cornerstone and adversely affected the cornerstone objective to ensure the
availability, reliability, and capability of systems that respond to initiating events to
prevent undesirable consequences. Using NRC Manual Chapter 0609,
Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the
finding was determined to be of very low safety significance because the finding
did not result in an actual loss of safety related equipment for greater than the
technical specification allowed outage time and did not represent a loss of
equipment designated as risk-significant in the maintenance rule. The finding did
not have a crosscutting aspect because it was not representative of current
licensee performance. (Section 4OA2.5b).
- Green. The team identified a Green noncited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure
of the licensee to follow the operability determination Procedure ODA-309,
Operability Determination and Functionality Assessment Program. Specifically,
the licensee did not appropriately evaluate a long-standing degraded condition
such that the diesel generators would remain operable for their mission time as
required by Procedure ODA-309. As a result, adequate compensatory measures
were not established to ensure operability. The licensee entered the finding into
the corrective action program as Condition Report CR 2011-008508.
The performance deficiency was determined to be more than minor because it
was associated with the equipment performance attribute of the Mitigating
System Cornerstone and affects the cornerstone objective to ensure the
availability and reliability of safety related diesel generators that respond to
initiating events to prevent undesirable consequences in that the safety related
diesel generators supply power to safety related loads. Because Manual
Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of
Findings, was not well suited for this finding, a Phase 3 Risk Significance
Estimation was required. A Region IV senior reactor analyst performed a
bounding Phase 3 significance determination and determined that the finding
was of very low safety significance. The bounding change to core damage
frequency was 6.7E-7/year. The simplified plant analysis risk (SPAR) model
does not include the contribution of the recently installed alternate power
generators, which would lower the risk significance of a safety related diesel
generator failure for the station blackout sequences, which comprise most of the
risk of this finding. The team determined that there was a crosscutting aspect in
the area of human performance decision-making because the licensee failed to
use conservative assumptions in decision making in the assessment of
operability H.1(b) (Section 4OA2.5c).
- Green. The team identified a Green noncited violation of 10 CFR Part 50,
Appendix B, Criterion XVI, Corrective Action, in that the licensee did not correct
a condition adverse to quality regarding the safety related diesel generators.
-4- Enclosure
Specifically, as of July 12, 2011, the licensee failed to assure that the identified
broken cam cover bolts on the diesel generators were adequately corrected. The
licensee entered the finding into the corrective action program as Condition
Report CR 2011-008505.
The performance deficiency was determined to be more than minor because it
was associated with the equipment performance attribute of the Mitigating
System Cornerstone and affects the cornerstone objective to ensure the
availability and reliability of safety related diesel generators that respond to
initiating events to prevent undesirable consequences in that the safety related
diesel generators supply power to vital and safety related loads. Because
Manual Chapter 0609, Attachment 4, Phase 1 Initial Screening and
Characterization of Findings, was not well suited for this finding a Phase 3 Risk
Significance Estimation was required. A Region IV senior reactor analyst
performed a bounding Phase 3 significance determination and found that the
finding was of very low safety significance. The bounding change to core
damage frequency was 6.7E-7/year. The SPAR model does not include the
contribution of the recently installed alternate power generators, which would
considerably lower the risk significance of safety related diesel generator failure
for the station blackout sequences, which comprise most of the risk of this
finding. The team determined that there was a crosscutting aspect in the area of
problem identification and resolution because the licensee failed to thoroughly
evaluate problems such that the resolutions address causes and extent of
conditions, as necessary P.1(c) (Section 4OA2.5d).
B. Licensee-Identified Violations
None
-5- Enclosure
REPORT DETAILS
4. OTHER ACTIVITIES (OA)
4OA2 Problem Identification and Resolution (71152)
The team based the following conclusions on the sample of corrective action documents
that were initiated in the assessment period, which ranged from August 15, 2009, to the
end of the on-site portion of this inspection on July 28, 2011.
.1 Assessment of the Corrective Action Program Effectiveness
a. Inspection Scope
The team reviewed documents, interviewed personnel, attended meetings, and walked
down plant equipment. The documents reviewed included over 450 corrective actions,
self-assessments, evaluations and station procedures including associated root cause,
apparent cause, and direct cause evaluations to determine if problems were being
properly identified, characterized, and entered into the corrective action program for
evaluation and resolution.
The team verified that the licensee entered problems into the corrective action program
for resolution. The team reviewed the details of the information related to the condition
reports to ensure that the evaluations were thorough and complete. The team reviewed
the licensees determinations on the extent of cause and condition for the problems, as
well as how the licensee assessed previous occurrences. The team assessed how the
licensee prioritized problems so that corrective actions were appropriate and timely. In
addition, the team verified the effectiveness of corrective actions, completed or planned,
and looked for additional examples of similar problems.
In order to accomplish the above, the team reviewed approximately 300 condition
reports out of approximately 26,000 that had been issued during the assessment period.
The team also reviewed a sample of system health reports, self-assessments, trending
reports and metrics, selected logs, audits, operability evaluations, and results from
surveillance tests and preventive maintenance tasks. The team reviewed a sample of
corrective actions closed to other corrective action documents.
The team reviewed a sample of system health reports, operability determinations, self-
assessments, trending reports and metrics, and various other documents related to the
corrective action program. The team evaluated the licensees efforts in establishing the
scope of problems by reviewing selected logs, work requests, self-assessment results,
audits, system health reports, action plans, and results from surveillance tests and
preventive maintenance tasks. The team reviewed work requests and attended the
licensees daily plan of the day meeting, a corrective action review board, a station
ownership committee, and a management review meeting to assess the reporting
threshold, prioritization efforts, and significance determination process, as well as
-6- Enclosure
observing the interfaces with the operability assessment and work control processes,
when applicable.
The team conducted interviews with plant personnel to identify other processes that may
exist where problems may be identified and addressed outside the corrective action
program.
The team reviewed corrective action documents that addressed past NRC-identified
violations to ensure that the corrective action addressed the issues as described in the
inspection reports. The team reviewed a sample of corrective actions closed to other
corrective action documents to ensure that corrective actions were appropriate and
timely.
The team considered risk insights from both the NRCs and the licensees risk
assessments to focus the sample selection and plant tours on risk significant systems
and components. The team selected the following risk significant systems:
- Safety related diesel generators
- Safety related service water system
- 480 volt electrical system
- Refueling water storage and condensate storage tanks
- Chemical and volume control system
The samples reviewed by the team focused on, but were not limited to, these systems.
The team also expanded their review to include five years of evaluations involving the
safety related diesel generators and service water systems to determine whether
problems were being effectively addressed. The team conducted a walkdown of these
systems to assess whether problems were identified and entered into the corrective
action program.
b. Assessments
1. Effectiveness of Problem Identification
The team concluded that the licensee identified conditions adverse to quality and
entered them into the corrective action program in accordance with the licensees
corrective action program guidance and NRC requirements. The team determined that
the licensee identified problems at a low threshold and entered them into the corrective
action program. However, the team identified problems during the team walkdown that
should have been previously recognized.
- Auxiliary feedwater cross connect valves were identified by team as leaking
grease. Condition Report CR 2011-007845 documents this issue.
-7- Enclosure
- Auxiliary feedwater inboard bearing oil level was identified as high out of the
normal band, requiring that excessive oil be drained. Condition Report CR 2011-
007851 documents this issue.
- A technically inadequate scaffolding procedure for seismic limitations of
scaffolding near safety related equipment. Condition Report CR 2011-007907
documents this issue.
The team did not identify any conditions adverse to quality that were not placed in the
corrective action program.
2. Effectiveness of Prioritization and Evaluation of Issues
The team concluded that, generally, the licensee effectively evaluated problems.
However, the team determined that there were several indications of weak evaluations
of long term problems.
- The team identified several instances where operability determinations on safety
related equipment did not properly consider the mission time which resulted in
acceptance of long term degraded conditions.
- The team also identified that there was a mindset that long term degraded
conditions were acceptable because there was no immediate impact to
operability.
- The team identified a work backlog in certain programs that were not being
properly addressed by key performance indicators.
3. Effectiveness of Corrective Action Program
The team concluded that actions to correct problems were generally effective.
However, the team identified three examples of conditions where corrective actions have
not been effective:
- Failures of diesel generator cam cover bolts, which were identified in 1995 but
replaced as they occurred.
- Reliability issues with the safety related service water vacuum breaker, such that
the vacuum breaker does not open when required. This has been an ongoing
issue since 2002.
- Numerous repeated failures of globe type (HermaValves) drain and vent valves
occurring since 2004. These were caused by yoke bushing failures and over-
torquing. The most recent failure resulted in an unusual event.
-8- Enclosure
.2 Assessment of the Use of Operating Experience
a. Inspection Scope
The team examined the licensees program for reviewing industry operating experience,
including reviewing the governing procedure and self-assessments. The team reviewed
a sample of industry operating experience evaluations to assess whether the licensee
had appropriately evaluated the notifications for relevance to the facility. The team also
reviewed assigned actions to ensure they were appropriate. The team reviewed a
sample of root and apparent cause evaluations to ensure that the licensee had
appropriately included industry operating experience.
b. Assessment
Overall, the team concluded that the licensee generally evaluated industry operating
experience for relevance to the facility, and appropriately entered applicable operating
experience into the corrective action program. The team concluded that operating
experience was appropriately included in causal evaluations. However, in two cases the
team determined that actions were not thorough enough regarding improper lubrication
of valve stem shafts on safety related motor operated valves.
- In April 8, 2011, during motor operated valve (MOV) testing in refueling outage
2RF12, Condition Report CR 2011-004136 documents Valve 2-8000A failure of
minimum thrust requirements in closed direction because of lack of lubricant.
- The team determined inadequate lubrication of motor operated valve stems,
which could be indicative of an inadequate procedure since full stroke and
inspection of stems were not possible. The licensees engineering staff
implemented limited actions documented in Condition Report CR 2007-002872,
USA-STARS self-assessments, which recommended full stroke for good
lubrication, but actions did not follow through to include the valves that could not
be fully inspected.
.3 Assessment of Self-Assessments and Audits
a. Inspection Scope
The team reviewed a sample of licensee self-assessments and audits to assess whether
the licensee was regularly identifying performance trends and effectively addressing
them. The team also reviewed audit reports to assess the effectiveness of assessments
in specific areas. The specific self-assessment documents and audits reviewed are
listed in the attachment.
The team also reviewed several licensee observations recorded by management to
ensure that issues were properly documented at the appropriate level. The team also
reviewed adverse trends documented in several areas including contamination events
occurring between the last two refueling outages.
-9- Enclosure
b. Assessment
The team concluded that the licensee had an effective self-assessment and audit
process. Licensee management was involved with developing tactical self-
assessments. The team determined self-assessments were self-critical and thorough
enough to identify deficiencies. The team noted that the licensee had improved their
operating experience program to ensure adequate overview by management and to
provide resources by assigning tactical and strategic related self-assessments.
Strategic self-assessments included personnel from outside organizations, and tactical
self-assessments received division management overview. The team noted the
licensee was reviewing actions to improve overdue self-assessments by improving
management oversight and effectiveness of the self-assessment review board.
However, the team also noted that the licensee divided industry operating events into
three separate organizations with an emphasis on processing third party significant
operating events. The team determined that the licensee limited self-assessments to
these significant operating events programs and did not include some of the engineering
related vendor document tracking programs. The team noted these programs had large
backlogs and were not included in key performance improvement indicators. These
programs included vendor document tracking reports (part of industry operating events
program), and preventive maintenance (PM) component basis feedback.
.4 Assessment of Safety-Conscious Work Environment
a. Inspection Scope
The team performed a review of the employee concern program known as SafeTeams
and conducted individual interviews of 28 licensees personnel. The interviewees
represented various functional organizations including radiation protection, operations,
maintenance, security, and supervision. Several plant activities were also observed
including Unit 1 plant startup and maintenance on a safety related diesel generator.
These interviews and observations were designed to elicit a qualitative assessment of
the degree to which the interviewees believed station management had established and
maintained a safety-conscious work environment.
In addition, the team reviewed the results of the licensees 2008 and 2010 Nuclear
Safety Culture Assessment results, as well as the licensees actions to address identified
concerns.
b. Assessment
Based on the results of the safety culture surveys and the focus groups, the team found
that the licensees programs had established a healthy safety-conscious work
environment in that every worker who had been interviewed by the team indicated they
felt free to raise safety concerns both to their management and to the NRC without fear
of retaliation. Workers felt comfortable using all avenues available to them in raising
- 10 - Enclosure
concerns that included writing condition reports, talking with their supervisors, informing
SafeTeam or management, and raising concerns with the NRC.
The team determined that individuals interviewed were collectively and individually
willing to raise nuclear safety concerns, knew of various ways to document concerns,
had not individually experienced retaliation for bringing up issues, and believed that the
licensees management generally supported employees raising nuclear safety concerns.
.5 Specific Issues Identified During This Inspection
a. Failure to Conclude a Change to the Final Safety Analysis Report Required Prior NRC
Review and Approval
Introduction. The team identified a Severity Level IV noncited violation of 10 CFR 50.59,
Changes, Tests, and Experiments, associated with the failure to conclude that a
change from the Final Safety Analysis Report did not require prior NRC review and
approval prior to implementation.
Description. The design function of the diesel generator jacket water cooling system is
to remove heat generated from operation of the safety related diesel generators under
transient and accident conditions including design basis accidents and loss-of-offsite
power. The licensee performed evaluation EV-2002-001666-02 that established a new
leakage rate of 2.4 gallons per hour as the maximum jacket water leakage rate to
maintain operability. This value was based on conditions where the jacket water level in
the standpipe was assumed to be at the low-level alarm set point with no operator
interaction for 7 days. This leakage rate was obtained from calculation ME-CA-0000-
5016, which had determined that the leakage rate of 1.5 gallons per hour, specified in
Final Safety Analysis Report Section 9.5.5.2, was a conservative acceptance criterion.
On March 13, 2004, the licensee performed an applicability screening, in accordance
with their 50.59 Resource Manual, prior to changing the acceptance criterion for
allowable jacket water leakage rate in the Final Safety Analysis Report, but incorrectly
concluded that an evaluation was not required for a change that involved manual
operator actions and a change to the allowable jacket water leak rate in support of a
design function that is credited in the Final Safety Analysis Report.
The licensee concluded that the proposed activity to increase the allowable leakage rate
did not involve a change to a structure, system, or component that adversely affected a
Final Safety Analysis Report described design function; as a result, the licensee did not
perform an evaluation to determine whether the proposed activity required NRC review
and approval prior to implementation. Subsequently, the licensee changed the Final
Safety Analysis Report from there is potentially 310 gallons of water available to replace
a leakage up to 1.5 gallons per hour for seven (7) days of continuous operation to there
is potentially 408 gallons of water available to replace a leakage up to 17 gallons per
hour for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of continuous operation. To change the leakage rate, the licensee
evaluation EV-CR 2004-000430-01-12, the licensee credited manual operator action to
- 11 - Enclosure
refill the jacket water system 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-accident to make-up the jacket water
system to further justify the new acceptance criterion.
Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes,
Tests, and Experiments, stated that Nuclear Energy Institute NEI 96-07, Guidelines for
10 CFR 50.59 Evaluations, provides methods that are acceptable to the NRC staff for
complying with the provisions of 10 CFR 50.59.
Using the guidance provided in NEI 96-07, the team determined that the proposed
activity, changing the acceptance criterion for allowable jacket water leakage in the Final
Safety Analysis Report, screened in because the activity affected a design function of a
structure, system, or component (i.e., the ability to remove heat from the diesel
generators during operation) by substituting manual action by the operators to make up
for increase jacket water leakage. Screening is performed by the licensee to determine
proposed activity should be evaluated against the criteria specified in
A 10 CFR 50.59 evaluation is required for changes that adversely affect design
functions. The team determined the change was adverse and required an evaluation
because the change involved the addition of manual operator action to refill the jacket
water system after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-accident. Further, the team concluded that the change
resulted in a more than minimal increase in the likelihood of occurrence of a malfunction
of a structure, system, or component important to safety because the change involved
substituting manual operation action to support a design function credited in the Final
Safety Analyses Report. Because the activity resulted in a more than minimal increase
in the likelihood of occurrence of a malfunction of a structure, system, or component
important to safety, the licensee must apply for and obtain a license amendment per 10 CFR 50.90 before implementing the activity. The licensee entered this issue into the
corrective action program as Condition Report CR 2011-008509.
Analysis. The failure of the licensee to adequately evaluate implementing a change to
the Final Safety Analysis Report concerning a change in acceptable jacket water leak
rate and addition of manual actions to refill the jacket water system after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
following a loss of offsite power, was contrary to 10 CFR 50.59(c)(2) and was a
performance deficiency. This finding was more than minor because there was a
reasonable likelihood that the change would require a prior NRC approval. Violations of
10 CFR 50.59 are violations that potentially impede or impact the regulatory process and
are processed through Traditional Enforcement. As required by Section 7.3 of the
Enforcement Policy, the team performed a Phase 1 screening in accordance with
Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization
of Findings, to determine the significance of the finding. The team determined that the
finding is of very low safety significance (Green) because the finding: (1) was not a
design or qualification issue confirmed not to result in a loss of operability or
functionality; (2) did not represent an actual loss of safety function of the system or train;
(3) did not result in the loss of one or more trains of nontechnical specification
equipment; and (4) did not screen as potentially risk significant due to a seismic,
flooding, or severe weather initiating event. Since violations of Title 10 CFR 50.59 may
- 12 - Enclosure
result in conditions evaluated as having very low safety significance by the Significance
Determination Process, the team categorized the finding as Severity Level IV in
accordance with the Enforcement Manual. The finding was a violation determined to be
of very low safety significance, was not repetitive or willful, and was entered into the
corrective action program. Therefore, this violation is being treated as a noncited
violation consistent with the NRC Enforcement Policy.
The performance deficiency is more than minor because there was a reasonable
likelihood that the change would require a prior NRC approval. Violations of 10 CFR 50.59 are violations that potentially impede or impact the regulatory process and are
processed through traditional enforcement. Violations of 10 CFR 50.59 are processed
through examples of Section 6.1 of the Enforcement Policy, and although the
significance determination process is not designed to assess the significance of
violations that potentially impact or impede the regulatory process, the staff has
determined that the significance of a 10 CFR 50.59 violation can be accessed through
the significance determination process. Therefore, the team performed a Phase 1
screening in accordance with NRC Manual Chapter 0609, Attachment 4, Phase 1 -
Initial Screening and Characterization of Findings, to determine the significance of the
finding. The team determined that the finding was of very low safety significance
(Green) because the finding: (1) was not a design or qualification issue confirmed not to
result in a loss of operability or functionality; (2) did not represent an actual loss of safety
function of the system or train; (3) did not result in the loss of one or more trains of
nontechnical specification equipment; and (4) did not screen as potentially risk
significant due to a seismic, flooding, or severe weather initiating event.
Since the violations of 10 CFR 50.59 resulted in a condition evaluated as having very
low safety significance by the significance determination process, the team categorized
the finding as Severity Level IV in accordance with the Enforcement Manual. The finding
was a violation determined to be of very low safety significance, was not repetitive or
willful, and was entered into the corrective action program. Therefore, this violation is
being treated as a noncited violation consistent with the NRC Enforcement Policy. The
team did not identify a crosscutting aspect with this finding since this performance issue
occurred in 2004 and is not reflective of current performance.
Enforcement. Title 10 CFR 50.59(c)(1), Changes, Tests, and Experiments, states, in
part, that a licensee may make changes in the facility as described in the Final Safety
Analysis Report (as updated) without obtaining a license amendment only if the change
does not result in more than a minimal increase in the likelihood of occurrence of a
malfunction of a structure, system, or component important to safety previously
evaluated in the Final Safety Analysis Report (as updated). Contrary to this
requirement, on September 28, 2004, the licensee made changes to the facility as
described in the final safety analysis report (as updated) without obtaining a license
amendment. Specifically, the licensee made changes to the acceptance allowable
diesel generator jacket water leakage in the Final Safety Analysis Report by substituting
manual operator action for increase jacket water leakage that resulted in more than a
minimal increase in the likelihood of occurrence of a malfunction of a structure, system,
- 13 - Enclosure
or component important to safety. Because this finding is of very low safety significance
and was entered into the corrective action program as Condition Report CR 2011-
008509, this violation is being treated as a noncited violation in accordance with Section
2.3.2 of the Enforcement Manual: NCV 05000445/2011006-01; 05000446/2011006-01,
Failure to Conclude a Change to the Final Safety Analysis Report Required Prior NRC
Review and Approval.
b. Inadequate Diesel Generator Jacket Water Fill Instructions
Introduction. The team identified a Green noncited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees
failure to have documented instructions for an activity affecting quality. Specifically, the
licensee did not have documented instructions for filling the diesel generator jacket water
system when the normal fill method would not be available during a loss of offsite power.
Description. On July 27, 2011, the team reviewed Procedure SOP-609A/B, Diesel
Generator System, Revision 12 and Procedure ALM-1301A/B, Alarm Procedure Diesel
Generator 1-01 Panel, Revision 5, to verify the compensatory measures that credited
operator action to fill the diesel generator jacket water system to compensate for system
leaks up to 17 gallons per hour. The team reviewed the diesel generator operating
series of procedures and the diesel generator alarm panel series of procedures and
identified that the procedures did not contain guidance on how to fill the jacket water
system during a condition where offsite power is not available. The procedures only had
guidance to fill the jacket water system using nonsafety-related equipment that did not
have an emergency power source. The team identified that the licensee had not
considered a scenario, in which offsite power would not be available to provide normal
makeup water and that alternative methods would be necessary. The team determined
that the licensee had implemented a change to the safety analysis in 2004, without an
adequate review of the design change. As a result of the teams questions, the licensee
documented this issue in Condition Report CR 2011-0008510.
Analysis. The licensees failure to have adequate instructions for filling the diesel
generator jacket water system was a performance deficiency. The performance
deficiency was more than minor because it was associated with the procedure quality
attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone
objective to ensure the availability, reliability, and capability of systems that respond to
initiating events to prevent undesirable consequences. Using NRC Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the
finding was determined to be of very low safety significance because the finding did not
result in an actual loss of safety related equipment for greater than the technical
specification allowed outage time and did not represent a loss of equipment designated
as risk-significant in the maintenance rule. The finding did not have a crosscutting
aspect because it was not representative of current licensee performance.
Enforcement. Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures,
and Drawings, requires, in part, that activities affecting quality shall be prescribed by
documented instructions of the type appropriate to the circumstances. Contrary to the
- 14 - Enclosure
above, as of July 27, 2011, the licensee failed to have adequate instructions for filling the
diesel generator jacket water system, an activity affecting quality, during a loss-of-offsite
power. Since the violation was of very low safety significance and was documented in
the licensees corrective action program as Condition Report CR 2011-008510, it is
being treated as a noncited violation, consistent with Section 2.3.2 of the NRC
Enforcement Policy: NCV 05000445/2011006-02; 05000446/2011006-02, Inadequate
Diesel Generator Jacket Water Fill Instructions.
c. Failure to follow Operability Determination Process for Degraded Diesel Generators
Introduction. The inspector identified a Green noncited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of the
licensee to follow Procedure ODA-309, Operability Determination and Functionality
Assessment Program. Specifically, the licensee did not appropriately evaluate a long-
standing degraded condition such that appropriate measures were taken to ensure the
safety related diesel generator would remain operable for the mission time, as required
by Procedure ODA-309.
Description. During interviews with licensee personnel and reviews of selected
operability determinations, the team noted a long-standing degraded condition with the
licensees diesel generators. The team was informed by the licensee that this condition
was well documented and was an industry wide issue.
On July 12, 2011, the team expressed the concern that the history of frequent cam cover
bolt failures could potentially impact the ability of safety related diesel generators to
perform their safety function for the mission time. The team identified that the licensee
failed to appropriately consider the impact to operability of this condition within the
mission time. Section 6.2.2 of Procedure ODA 309 states, in part, that if conditions
impact or potentially impact the ability of the technical specification structure, system or
component to perform its required function for the credited time duration (mission time),
then measures are needed to ensure the component will remain operable to provide the
specified safety function with the degraded or nonconforming condition for the required
mission time. In addition, Section 6.2.2 of Procedure ODA 309 also requires that
appropriate actions be taken if compensatory measures are required to maintain
operability. The concerns of the team regarding cam cover bolts failure rates were
documented on Condition Report CR 2011-007850.
The team reviewed Smart Form Technical Evaluation TE 95-0030 dated March 3, 1995,
which was used to justify operability. The evaluation provided an analysis of the safety
related diesel generator cam cover bolts and stated the minimum number of bolts
required to maintain the cam cover joint seal was based upon the maximum loading on
the cover plate. The licensee evaluation concluded that the cover plate joint was
acceptable and the safety related diesel generator would remain operable as long as at
least five bolts remained intact on each of the top or bottom of the covers. However, the
licensee did not consider the failure rate in the operability determination. The team
determined that the licensee did not use conservative decision making since the cam
cover bolts failures were a long standing condition and could be replaced when found
- 15 - Enclosure
broken. As a result, the licensee did not appropriately consider the mission time to
support the design function of safety related diesel generator in the operability
determination.
As a result of the teams questions, the licensee completed a new operability
determination and determined that the condition required compensatory measures for
the safety related diesel generator to remain operable during the mission time including
replacing bolts while a diesel generator was running. In addition, procedures were
changed to provide operators with instructions to specifically look for and identify bolt
failures while the diesel generators are running. The licensee also replaced all existing
safety related diesel generator cam cover bolts. The team observed this activity and
verified that replacement of cam cover bolts on an operating diesel was plausible.
Analysis. The failure to perform an adequate operability determination on the safety
related diesel generator was a performance deficiency. This finding was more than
minor because it was associated with the equipment performance attribute of the
Mitigating System Cornerstone and affected the cornerstone objective to ensure the
availability and reliability of safety related diesel generators that respond to initiating
events to prevent undesirable consequences in that the safety related diesel generators
supply power to vital and safety related loads. Specifically, the operability determination
did not ensure that the safety related diesel generators would remain operable for their
mission time and perform their safety function.
The Manual Chapter 0609, Attachment 4, Phase 1 Initial Screening, and
Characterization of Findings, was not well suited for this finding and a Phase 3 Risk
Significance Estimation was required. A Region IV senior reactor analyst performed a
bounding Phase 3 significance determination. The analyst estimated a bounding
change in core damage frequency (delta-CDF) for the performance deficiency using the
following assumptions:
- Based on calculations performed by the licensee, a total of 5 bolt failures on the
top or the bottom of the cam cover were sufficient to cause failure of the diesel
generator.
- The exposure period was one year.
- A bolt failure history over the past five years was used and a bounding
assumption made that all failures are assumed to be on the top of one of the 4
cam covers on each safety related diesel generator. Therefore, any safety
related diesel generator start that had 5 bolt failures was assumed to cause a
safety related diesel generator failure.
- The safety related diesel generator recovery following failure from a cam cover
failure was assumed to follow the nominal recovery probabilities. This
assumption would be important only for cutsets where both safety related diesel
- 16 - Enclosure
- generators fail to run from the bolt failures and represented a small portion of the
delta-CDF. Otherwise, the other safety related diesel generator was available for
recovery.
Data was reviewed over a 5-year period for all four safety related diesel generators at
the site. There were no cases where more than 3 bolts failed during a single safety
related diesel generator run. During this time, the safety related diesel generators were
run an estimated 300 times collectively with no failures of the cam cover seals. As a
bounding assumption, the analyst assumed that the probability that a safety related
diesel generator would fail from a cam cover failure is 1/150 or 6.7E-3 (equivalent to 2
failures in 300 runs).
Most safety related diesel generator runs are for less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; therefore, it was
likely that additional bolt failures would occur in the 24-hour period following an actual
event than were reported in the data. However, whenever the safety related diesel
generators were running, they were inspected continuously, and it was demonstrated
that bolt failures can be detected and the bolts can be replaced while the safety related
diesel generators continued to run. The analyst concluded that these two effects cancel
out and that the bolt failure data was representative of the expected 24-hour
performance. The probability of the fail-to-run basic event for both safety related diesel
generators was increased by 6.7E-3 in the Comanche Peak simplified plant analysis risk
(SPAR) model, Revision 8.15; the model was run at a truncation of 1.0E-12, with
average test and maintenance. The result of the Phase 3 bounding analysis resulted in
a delta-CDF of 6.7E-7/year, indicating this finding is of very low safety significance
(Green).
The SPAR model included the contribution of the recently installed alternate power
generators, which lowers significantly the risk significance of a safety related diesel
generator failure for the station blackout sequences, which comprise the majority of the
risk of this finding. The team determined that there was a cross cutting aspect in the
area of human performance decision-making because the licensee failed to use
conservative assumptions in decision making in the assessment of operability H.1(b).
Enforcement. Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures,
and Drawings, requires, in part, that activities affecting quality shall be prescribed by
documented instructions, procedures or drawings, of a type appropriate to the
circumstances and shall be accomplished in accordance with these instructions,
procedures or drawings. Contrary to the above, prior to July 12, 2011, the licensee did
not adequately implement the requirements of operability determination process in
accordance with Procedure ODA 309. Specifically, Section 6.2.2 of ODA 309 requires
the licensee to assess degraded and nonconforming conditions to consider mission time
and establish compensatory measures as interim actions to maintain, enhance, or
restore operability of safety-related equipment until final corrective actions have been
completed. Because the finding is of very low safety significance and has been entered
into the licensees corrective action program as Condition Reports CR 2011-007850 and
CR 2011-008508, this violation is being treated as a noncited violation consistent with
- 17 - Enclosure
Section 2.3.2 of the NRC Enforcement Policy: NCV 05000445/2011006-03;
05000446/2011006-03, Failure to follow Operability Determination Process for
Degraded Diesel Generators.
d. Repeated Diesel Generator Cam Cover Bolt Failures
Introduction. The team identified a Green noncited violation of 10 CFR Part 50,
Appendix B, Criterion XVI, Corrective Action, in that the licensee did not correct a
condition adverse to quality regarding the safety related diesel generators.
Description. During reviews of the corrective actions for selected safety related systems,
the team noted several condition reports that documented repeated failures of
components such as cam cover bolt failures, after running the safety related diesel
generators. The safety related diesel generators each have eight large cam covers
fastened by twenty-six bolts around a covers perimeter. Operability of the safety related
diesel generator may be impacted if these covers become loose since cam covers are
required to ensure a mechanical pressure boundary so that a slight vacuum is
maintained on the crankcase. The cam covers also function to keep air out and oil in to
prevent a potentially hazardous combustible mixture. The configuration of the bolts was
nine bolts across the top side of the cam cover, nine bolts across the bottom row, and
four bolts on each of the shorter sides.
Procedure STA-422, Processing Condition Reports, Revision 25, describes a level C
condition, as a condition that involves minimal impact on safe reliable plant operation
and is of low safety significance that an apparent cause determination is not required.
The team determined that, although the licensee had identified each instance of a cam
cover bolt failure, the condition adverse to quality was not corrected in a timely manner,
as made evident by the recurrence and the failure to evaluate the condition adverse to
quality in accordance with the site corrective action process. The team determined that
cam bolt failures, based on the number and repeated nature of the issue, should have
been classified as a higher condition report and should not have continued to have been
treated as having minimum impact to safe reliable operation or that an apparent cause
determination was not required. The team noted that all cam cover bolts failure
conditions had been assigned C or D level condition reports. Further, the team
determined Procedure STA-422 prescribed that an apparent cause of the issue be
documented (if categorized a Category B condition, which is a higher category), and
corrective actions taken to correct the condition and to address the apparent cause(s).
The team noted that an apparent cause evaluation had not been performed.
In 1995, because of repeated bolt failures, the licensee concluded that the cover plate
joint was acceptable and that the safety related diesel generators would remain operable
with five bolts intact on each of the top or bottom of each cover. Following routine
operations of the safety related diesel generators, the licensee replaced the broken bolts
when found. However, the licensee failed to consider the recurring aspect of the
problem and its effect on functionality. As a result, the licensee failed to thoroughly
evaluate and address the impact to operability associated with the potential for cam
cover bolt failures over a long period of operation.
- 18 - Enclosure
Based on the number of cam cover bolts failures identified by licensee since 1995 to
July 2011, the team determined that the licensee did not implement actions to correct
the repeated failures in accordance with Procedure STA-422. As a result, there were
additional occurrences of cam bolt failures from the operation of the safety related diesel
generators, that could have impacted the safety related diesel generators operability.
Analysis. The team determined that the failure to implement corrective action for the
cam cover bolt failures was a performance deficiency. This finding was more than minor
because, if left uncorrected, the performance deficiency would have the potential to lead
to a more significant concern. The NRC Manual Chapter 0609, Attachment 4, Phase 1
Initial Screening and Characterization of Findings, was not well suited for this finding
and it required a Phase 3 Risk Significance Estimation. A Region IV senior reactor
analyst performed a bounding Phase 3 significance determination. The analyst
estimated a bounding change in core damage frequency (delta-CDF) for the
performance deficiency using the following assumptions:
- Based on calculations performed by the licensee, a total of 5 bolt failures on the
top or the bottom of a cam cover were sufficient to cause failure of a safety
related diesel generator.
- The exposure period was one year.
- A bolt failure history over the past five years was used and a bounding
assumption made that all failures are assumed to be on the top of one of the 4
cam covers on each safety related diesel generator. Therefore, any safety
related diesel generator start that had 5 bolt failures was assumed to cause a
safety related diesel generator failure.
- The safety related diesel generator recovery following failure from a cam cover
failure was assumed to follow the nominal recovery probabilities. This
assumption would be important only for cutsets where both safety related diesel
generators fail to run from the bolt failures and represented a small portion of the
delta-CDF. Otherwise, the other safety related diesel generator was available for
recovery from some other problem.
Data was reviewed over a 5-year period for all four safety related diesel generators at
the site. There were no cases where more than 3 bolts failed during a single safety
related diesel generator run. During this time, the safety related diesel generators were
run an estimated 300 times collectively with no failures of the cam covers. As a
bounding assumption, the analyst assumed that the probability that a safety related
diesel generator would fail from a cam cover failure is 1/150 or 6.7E-3 (equivalent to 2
failures in 300 runs).
Most safety related diesel generator runs are for less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; therefore, it was
likely that additional bolt failures would occur in the 24-hour period following an actual
event than were reported in the data. However, whenever the safety related diesel
generators were running, they were inspected continuously, and it was demonstrated
- 19 - Enclosure
that bolt failures can be detected and the bolts can be replaced while an safety related
diesel generator continued to run. The analyst concluded that these two effects cancel
out and that the bolt failure data was representative of the expected 24-hour
performance. The probability of the fail-to-run basic event for both safety related diesel
generators was increased by 6.7E-3 in the Comanche Peak SPAR model, the model
was run at a truncation of 1.0E-12, with average test and maintenance. The result of the
Phase 3 bounding analysis resulted in a delta-CDF of 6.7E-7/year, indicating this finding
is of very low safety significance (Green).
The SPAR model included the contribution of the recently installed alternate power
generators, which lowers significantly the risk significance of a safety related diesel
generator failure for the station blackout sequences, which comprise the majority of the
risk of this finding.
The team determined that there was a cross cutting aspect in the area of problem
identification and resolution because the licensee failed to thoroughly evaluate problems
such that the resolutions address causes and extent of conditions, as necessary P.1(c).
Enforcement. The team identified a Green noncited violation of 10 CFR Part 50,
Appendix B, Criterion XVI, Corrective Action, which states, 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. Contrary to the above, the
licensee failed to assure that conditions adverse to quality were promptly identified and
corrected. Specifically, as of July 12, 2011, the licensee failed to assure that the
identified broken cam cover bolts on the safety related diesel generators were effectively
corrected. This finding was entered into the licensees corrective action program as
Condition Report 2011-008505. Because this violation was of very low safety
significance (Green) and has been entered into the licensees corrective action program,
this violation is being treated as a noncited violation, consistent with the NRC
Enforcement Policy: NCV 05000445/2011006-04; 05000446/2011006-04, Repeated
Diesel Generator Cam Cover Bolt Failures.
4OA6 Meetings
Exit Meeting Summary
On July 28, 2011, the team presented the inspection results to Mitch L. Lucas, Site Vice
President, and other members of the licensee staff. The licensee management acknowledged
the issues presented. The team asked the licensee management if there were any materials in
the procession of the team, which should be considered proprietary. No proprietary information
was identified.
4OA7 Licensee-Identified Violations
None
- 20 - Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
D. Ambrose, Manager Corrective Action Program
J. Audas, Manager SafeTeam
C. Beerck, Senior Nuclear Auditor
C. Cummins, Supervisor Performance Improvement
D. Fuller, Manager Emergency Planning
T. Gibbs, SafeTeam Supervisor
T. Gilder, Director Performance Improvement
D. Goodwin, Director Engineering Support
J. Henderson, Manger Engineering Smart Team
M. Lucas, Site Vice President
M. Marler, Director Organization Effectiveness
G. Merka, Regulatory Affairs Engineer
C. Miller, Manager Plant Reliability
K. Nickerson, Director Site Engineering
B. Patrick, Director Maintenance
J. Patton, Manager Quality Assurance
W. Reppa, Manager System Engineering
J. Seawright, Regulatory Affairs Engineer
S. Sewell, Director Nuclear Operations
T. Terryah, Manager Engineering Smart Team
T. Tigner, Corrective Action Supervisor
D. Weyandt, Senior System Engineer
L. Zimmerman, Manager Procurement Engineering and Programs
NRC Personnel
B. Tindell, Resident Inspector
M. Hay, Acting Chief PSB-1
A1-1 Attachment
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
Failure to Conclude a Change to the Final Safety Analysis05000445/2011006-01
NCV Report Required Prior NRC Review and Approval
(Section 4OA2.5a)05000445/2011006-02 Inadequate Diesel Generator Jacket Water Fill Instructions
NCV 05000446/2011006-02 (Section 4OA2.5b)05000445/2011006-03 Failure to Follow Operability Determination Process for
NCV 05000446/2011006-03 Degraded Safety Related Diesel Generators (Section 4OA2.5c)05000445/2011006-04 Repeated Diesel Generator Cam Cover Bolt Failures
NCV 05000446/2011006-04 (Section 4OA2.5d)
LIST OF DOCUMENTS REVIEWED
Section 4OA2: Identification and Resolution of Problems
PROCEDURES
NUMBER TITLE REVISION / DATE
STA-213 Correspondence with Regulatory Agencies and Industry 3
Groups
ODA-104 Operations Department Document Control 14
STA-201 Procedure Use and Adherence 16
STA-424 Self-Assessment and Benchmarking Programs 5
NMG-130 CPNPP Observation Program July 7, 2011
ECE-5.01-04 Technical Evaluation of Replacement Items (TERI) November 9, 2009
SEC-108 Security Field Report, Condition Report, and Security 10
Reporting Requirements
ASG-025 Safeguards Information Records Processing Guideline 8
STA-308 Protection of Safeguards Information and Safeguards 13
Information - Modified
MSM-C0-3345 Emergency Diesel Engine Crankcase Relief Valve Inspection 2
SOP-609A Diesel Generator System 20
SOP-609B Diesel Generator System 12
ALM-1301A Alarm Procedure Diesel Generator 1-01 Panel 5
ALM-1302A Alarm Procedure Diesel Generator 1-02 Panel 5
ALM-1301B Alarm Procedure Diesel Generator 2-01 Panel 3
A1-2 Attachment
ALM-1302B Alarm Procedure Diesel Generator 2-02 Panel 2
NQA-2.08 Nuclear Industry Cooperative Audits 13
STA-114 Employee Concerns and Employee Protection
NQA-3.02 Audit and Surveillance Programs 6
EVAL-2011-003 Emergency Planning Changes January 19, 2011
QA 20110105 QA Surveillance Follow-up January 5, 2011
SA-2010-027 Self-Assessment of Flow Accelerated Corrosion Program February 5, 2010
(CR-2010-001181)
WCI-607 Fluid Leak Management Process 2
NMG-705 Corrective Action Review Board (CARB) Process February 17, 2011
Roll-up Meeting Center Of Excellence Periodic Report February 22, 2011
Summary
ODA-309 Operability Determination and Functionality Assessment 2
Program
STA-421 Initiation of Condition Reports 17
STA-426 Industry Operating Experience Programs 8
STA-602 Temporary Modifications and Transient Equipment 16
Placements
STA-744 Maintenance Effectiveness Monitoring Program 5
CP-201100923 Two Year Rolling Audit Schedule 6
STA-705 Radioactive Systems Leakage Inspection Program 6
STA-206 Review of Vendor Documents and Vendor Technical 24
Manuals
STA-422 Processing Condition Reports 25
DRAWINGS
NUMBER TITLE REVISION
M1-1900 Penetration Seal Schedule CP-7
WORK ORDERS
NUMBER
4078008 4077949
A1-3 Attachment
INFORMATION NOTICES
NUMBER TITLE DATE
Spurious Relay Actuations Result in Loss of Power to
2009-016 September 15, 2009
Safeguard Buses
2009-21 Incomplete Medical Testing for Licensed Operators September 30, 2009
Recent Human Performance Issues at Nuclear Power
2009-22 October 2, 2009
Plants
Degradation of Wire Rope Used in Fuel Handling
2009-20 October 7, 2009
Applications
2009-23 Nuclear Fuel Thermal Conductivity Degradation October 8, 2009
Sources of Information Related to Potential Cyber
2009-24 October 13, 2009
Security Vulnerabilities
Degradation of Neutron-Absorbing Materials in the
2009-26 October 28, 2009
Spent Fuel Pool
Failures of Motor operated Valves due to Degraded
2010-03 February 2, 2010
Stem Lubricant
Management of Steam Generator Loose Parts and
2010-05 February 3, 2010
Automated Eddy Current Data Analysis
2010-07 Welding Defects in Replacement Steam Generator April 5, 2010
Importance of Understanding Circuit Breaker Control
2010-09 April 14, 2010
Power Indications
Implementation of a Digital Control System Under 10
2010-10 May 28, 2010
CFR 50.59
2010-12 Containment Liner Corrosion June 18, 2010
2010-25 Inadequate Electrical Connections November 17, 2010
OBSERVATIONS
NUMBER TITLE DATE
2011-2929 Material Condition; Manager and Supervisors April 6, 2011
2011-3710 Eng04, Design Changes; Engineering April 28, 2011
2011-4195 OMOP 11.3 Pre-Job Briefs Operations April 24, 2011
2011-4229 Maintenance Interdepartmental Pre-Job Briefs April 26, 2011
2011-4892 Team Observation Engineering June 6, 2011
CONDITION REPORTS
2009-003470 2009-003479 2009-003786 2009-004311 2009-004817
2009-004819 2009-0054001 2009-005427 2009-005430 2009-005501
2009-005772 2009-005777 2009-006488 2009-008039 2009-008232
2010-001224 2009-001225 2009-004568 2010-004571 2010-005812
2010-006271 2010-010728 2010-007951 2010-009018 2009-004085
2010-011152 2010-003476 2011-002702 2011-004536 2011-007178
A1-4 Attachment
2010-009073 2010-004158 2011-006820 2011-007175 2011-007547
2011-007172 2011-006788 2009-003728 2011-007411 2011-005454
2011-001876 2011-007233 2011-007144 2011-005914 2010-009909
2010-006526 2011-007703 2011-005920 2010-000624 2009-008027
2010-010033 2009-08129 2010-010781 2010-007472 2011-002716
2010-010018 2010-005838 2010-009993 2010-009694 2010-007091
2010-003758 2010-010034 2009-008622 2010-010818 2009-002422
2010-005843 2011-003316 2011-008210 2011-007851 2010-005784
2010-003685 2010-003680 2010-004031 2010-002357 2010-003757
2010-007798 2010-008108 2010-001312 2011-008376 2010-004050
2011-000824 2009-002038 2010-001179 2009-001399 2011-000816
2003-001068 2011-002944 2010-010609 2010-006120 2010-008926
2010-008429 2011-000825 2011-000680 2011-000678 2010-000197
2010-008411 2009-005424 2010-001224 2011-004136 2010-002525
2010-005924 2011-007595 2011-008439 2007-000519 2010-005628
2010-001736 2011-007736 2011-007644 2011-007356 2010-005563
2011-003633 2010-005941 2010-006561 2010-003458 2010-010391
2009-006582 2010-001242 2010-010781 2010-006349 2011-002349
2010-004562 2010-009417 2010-003775 2010-003783 2010-008489
2010-002671 2009-006625 2009-008593 2010-003789 2010-009498
2009-005503 2010-008049 2010-011513 2010-002626 2011-002571
2010-006450 2009-004046 2009-004052 2009-004807 2011-004741
2011-002411 2009-006047 2010-001119 2010-004044 2011-004621
2010-003305 2010-007472 2010-010652 2010-011270 2011-004909
2010-006595 2011-004098 2011-001876 2010-011513 2011-005637
2009-001069 2009-008643 2010-000266 2010-000638 2011-005646
2009-004453 2010-003763 2010-006120 2010-006268 2011-005819
2009-005542 2009-000104 2009-000848 2009-000926 2009-001548
2009-002876 2009-004054 2009-005501 2009-004454 2009-005275
2009-006665 2010-000897 2011-005867 2009-004455 2004-000193
2010-005963 2010-004331 2009-006696 2010-007458 2007-001273
2010-002524 2010-001070 2011-003265 2011-000041 2011-001468
MISCELLANEOUS
NUMBER TITLE REVISION / DATE
ET31CQT091 Effective Issue Documentation Training March 2, 2009
ET31CQT101 STA 206 Rev 23 VETIP Procedure Training October 14, 2010
Operations Operator Burdens and Work - Arounds March 4, 2011
Guideline 36
EV-2009-06 Audit on Licensing, Permits and Reporting May 27, 2010
Assessment Plans- Loss Prevention Performance 6
Improvement Process
A1-5 Attachment
VL 07-001807 Letter from John Crane Company on Seal Information October 5, 2007
(CPES 200701528)
CR-2010-011152 Mid-Cycle Strategic Self-Assessment February 24, 2011
QA 20100913 Tactical Self-Assessment CARB actions not corrected September 13, 2010
EV-2010-007593 Tactical Self-Assessment Industry OE program August 10, 2010
EV-2010-008110 Assessment of Level C- Vulnerabilities in Programs for December 16, 2010
Assessing 10CFR21 Reports (QA 20100527)
CPES-M-2012 Piping and Equipment Insulation 9
DBD-ME-229 Component Cooling Water System 36
NEI 97-07 Guidelines for 10 CFR 50.59 Evaluations 3
STARS-DTI-001 Desk Top Instruction for the Avoidance of Substantive 0
Cross-Cutting Issues (SCCIs)
TB-09-4 Impact of Auxiliary Pump Heat on Westinghouse and 1
Combustion Engineering Analyses/Methodologies
ME(B)-240 Condensate Storage Tank Tech. Spec. Limit 4
TDM-804A CPSES: Technical Data Manual, Equipment Data: Tank 2
Height vs. Volume, Unit 1
TDM-804B CPSES: Technical Data Manual, Equipment Data: Tank 2
Height vs. Volume, Unit 2
EV-CR-2011- U2 DG Droop; Failure Analysis of L2-Auxiliary Switch
004184-3
EVAL-2010-006 Performance Indicator Process and Change February 8, 2010
Management
EV-CR-2010- The NRC CDBI team identified a potential violation of
006120-1 Technical Specification SR 3.3.4.2.
SA-CR-2010- Conduct a Strategic Self-Assessment of the Maintenance
011154 Rule Program
SA-CR-2010- This condition report will document the planning,
010589 performance, and reporting of a Strategic Self-
Assessment
EV-CR-2010- NRC IN2010-20: Turbine Driven Auxiliary Feedwater
008926-1 Pump Repetitive Failures
EV-CR-2009- The U.S. Nuclear Regulatory Commission (NRC) is
005424-01 issuing this information notice (IN) to alert addresses of
an event at Point Beach Nuclear Plant (PBNP) in which
spurious relay actuation resulted in the loss of offsite
power to a safety-related 480 Volts alternating current
(Vac) safeguards bus for more than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
A1-6 Attachment
EV-CR-2011- Valve 2-8000A (pressurizer block valve) failed to meet one
004136-1, 2, 4 of two requirements for the minimum thrust requirement in
the closed direction
NUREG/CR- Performance of MOV Stem Lubricants at Elevated October 2001
6750 Temperatures
EPG-03 Good Practice Engineering Program Guide: Motor-
Operated Valves
SA-2007-021 Motor Operated Valve Program
EVAL-2007-002 Equipment Reliability February 15, 2007
EV-CR-2009- The SSW Discharge Valves (u-HV-4286 and u-HV-4287)
003728 are Fisher style 9200 Series which use an elastomer seat
ring (not a full body liner) held in place by a bolted
retaining ring.
EV-CR-2010- Since Surveillance Work Order 3774545 was not
005913 performed by 02/11/2011, the purpose of this QTE is to
establish if 1CT-0309 is OPERABLE.
EV-CR-2011- Since the surveillance was missed on 2-HV-4572 during
003308 2RF11, the purpose of this QTE is to evaluate the risk
associated with the missed surveillance and document
further the OPERABILITY of 2-HV-4572.
EV-CR-2011- Received CCP 2 L\O CLR SSW RET FLO LO alarm on 1-
003605 ALB-1 window 2.11.
EV-CR-2006- On the 30th of March Unit 1 responded to an alarm on 1-
001255 ALB-1 window 2.12 for SIP 2 L/O Clr SSW Ret Flo Lo
Alarm.
REI-701 SSW Water Hammer Test 0
EV-CR-2011- During the performance of OPT-207B for Station Service
007598-22 Water Pump 2-02 (CP2-SWAPSW-02), vacuum breaker
CP2-SWVAVB-04 did not break vacuum during Step
8.2.R. Reference CR-2010-008411.
DBD-ME-233 Station Service Water System 21
EV-CR-2002- During surveillance testing on CP1-SWVAVB-04, vacuum
003545-8 breaker failed to lift at desired set pressure range of 0.001
to 0.249 PSIV.
EV-CR-2002- QTE-2002-3545-01 provides the technical justification for
003545-1 SSW Operability with the Vacuum Breakers on the system
stuck shut.
EV-CR-2010- Safety chiller 2-06 tripped on compressor high discharge
005628-4 temperature when all loads were removed from the chiller.
A1-7 Attachment
EV-CR-2010- At or around 8:10 am on 2/16/10 an employee was using
001588-24 a forklift in an attempt to load a Westinghouse training
component onto a truck for transport While lifting the
load, it fell sideways from the forklift and struck the driver
of the truck. The driver received a cut to his right ear.
EV-CR-2010- On March 4, 2010, the Gantry crane was being used to
002235-4 move Unit 1 Turbine Auxiliary Lube Oil Motor 1-B for
replacement. During the move, the crane came in contact
with scaffolding. The contact with the scaffolding resulted
in some minor damage to surrounding equipment
including insulation, a sight glass, and a bent handwheel
on 1-HD-0960. The control room received a momentary
MSR high level alarm.
EV-CR-2010- On April 5, 2010, the performance Unit 2 Train B Diesel
003305-18 Generator Integrated Test Sequence (ITS) Surveillance
on 10/12/2009, the diesel did not shift to isochronous
mode in the Loss of Offsite Power (LOOP) portion of the
test.
EV-CR-2010- On April 12, 2010 at approximately 0720, CPNPP
003783-7 experienced a line to insulator flashover between Startup
Transformer XST1 and the 138 kV switchyard while Unit 2
was operating at 100% power.
EV-CR-2010- On April 12, 2010, the CPNPP Operations Training
004194-3 Supervisor - Initial was notified that two of the nine
applicants from License Class 18 had failed the simulator
portion of the NRC operating license test.
EV-CR-2010- On March 17, 2011, during the response to NRC IN 2010-
006268-22 11, Design Basis Engineering determined that neither
Residual Heat Removal (RHR) train was capable of
performing the required Emergency Core Cooling System
(ECCS) injection function during at least three of the last
four outages (2RF10, 1RF13, and 2RF11).
EV-CR-2011- On January 13, 2011, Shift Operations discovered relay
000356-8 27-1/1A1 PROTECTIVE RELAY FOR REACTOR TRIP
severely chattering.
EV-CR-2011- During the February 2011 Triennial Fire Inspection,
001742-2 CPNPP received a Green Cited Violation due to a repeat
problem from the 2008 Triennial Fire Inspection for failing
to verify Station Service Water (SSW) flow to an operating
Diesel Generator within the required time line.
A1-8 Attachment
EV-CR-2009- CST: Calculation ME-CA-00005295 (Comanche Peak
004885-00 Unit 1 Minimum CST Volume for RSG/Uprate) does not
appear to account for leakage from the Auxiliary
Feedwater pump seals.
EV-CR-2010- Perform and Document Root Cause Analysis
010781-5
OE 26026 Painting Activities and Cleaning Agents Render August 17, 2009
Emergency Diesel Generators inoperable
A1-9 Attachment
Information Request
May 18, 2011, 2010
Biennial Problem Identification and Resolution Inspection - Comanche Peak
Inspection Report 2011006
This inspection will cover the period from July 27, 2009 to July 29, 2011. All requested
information should be limited to this period unless otherwise specified. To the extent possible,
the requested information should be provided electronically in Adobe PDF or Microsoft Office
format. Lists of documents should be provided in Microsoft Excel or a similar sortable format.
A supplemental information request will likely be sent during the week of June 6, 2011.
Please provide the following no later than June 3, 2011.
1. Document Lists
Note: for these summary lists, please include the document/reference number, the document
title or a description of the issue, initiation date, and current status. Please include long text
descriptions of the issues.
a. Summary list of all corrective action documents related to significant conditions
adverse to quality that were opened, closed, or evaluated during the period
b. Summary list of all corrective action documents related to conditions adverse to
quality that were opened or closed during the period
c. Summary lists of all corrective action documents which were upgraded or
downgraded in priority/significance during the period
d. Summary list of all corrective action documents that subsume or roll up one or
more smaller issues for the period
e. Summary lists of operator workarounds, engineering review requests and/or
operability evaluations, temporary modifications, and control room and safety
system deficiencies opened, closed, or evaluated during the period
f. Summary list of plant safety issues raised or addressed by the Employee
Concerns Program (or equivalent)
g. Summary list of all Apparent Cause Evaluations completed during the period
h. Summary list of all Root Cause Evaluations planned or in progress but not
complete at the end of the period
2. Full Documents, with Attachments
a. Root Cause Evaluations completed during the period
b. Quality assurance audits performed during the period
A2-1 Attachment 2
c. All audits/surveillances performed during the period of the Corrective Action
Program, of individual corrective actions, and of cause evaluations
d. Corrective action activity reports, functional area self-assessments, and non-
NRC third party assessments completed during the period (do not include INPO
assessments)
e. Corrective action documents generated during the period for the following:
- NCVs and Violations issued to Comanche Peak
- LERs issued by Comanche Peak
f. Corrective action documents generated for the following, if they were determined
to be applicable to Comanche Peak (for those that were evaluated but
determined not to be applicable, provide a summary list):
- NRC Information Notices, Bulletins, and Generic Letters issued or
evaluated during the period
- Part 21 reports issued or evaluated during the period
- Vendor safety information letters (or equivalent) issued or evaluated
during the period
- Other external events and/or Operating Experience evaluated for
applicability during the period
g. Corrective action documents generated for the following:
- Emergency planning drills and tabletop exercises performed during the
period
- Maintenance preventable functional failures which occurred or were
evaluated during the period
- Adverse trends in equipment, processes, procedures, or programs which
were evaluated during the period
- Action items generated or addressed by plant safety review committees
during the period
3. Logs and Reports
a. Corrective action performance trending/tracking information generated during the
period and broken down by functional organization
b. Corrective action effectiveness review reports generated during the period
A2-2 Attachment 2
c. Current system health reports or similar information
d. Radiation protection event logs during the period
e. Security event logs and security incidents during the period (sensitive information
can be provided by hard copy during first week on site)
f. Employee Concern Program (or equivalent) logs (sensitive information can be
provided by hard copy during first week on site)
g. List of Training deficiencies, requests for training improvements, and simulator
deficiencies for the period
4. Procedures
a. Corrective action program procedures, to include initiation and evaluation
procedures, operability determination procedures, apparent and root cause
evaluation/determination procedures, and any other procedures which implement
the corrective action program at Comanche Peak.
b. Quality Assurance program procedures
c. Employee Concerns Program (or equivalent) procedures
d. Procedures which implement/maintain a Safety Conscious Work Environment
5. Other
a. List of risk significant components and systems
b. Organization charts for plant staff and long-term/permanent contractors
c. List of Corrective actions documented between April 2006 - April 2011
associated with the following risk significant systems:
- Safety Related Service Water System
Note: Corrective action documents refers to condition reports, notifications, action requests,
cause evaluations, and/or other similar documents, as applicable to Comanche Peak.
A2-3 Attachment 2
As it becomes available this information should be uploaded on the Certrec IMS website. When
these documents have been compiled but no later than June 3, 2011, please download these
documents onto a CD and sent it via overnight carrier to:
U.S. NRC Region IV
612 E. Lamar Blvd.
Suite 400
Arlington, TX 76011
Please note that the NRC is not able to accept electronic documents on thumb drives or other
similar digital media. However, CDs and DVDs are acceptable.
A2-4 Attachment 2