ML093060038
ML093060038 | |
Person / Time | |
---|---|
Site: | Harris |
Issue date: | 10/30/2009 |
From: | Daniel Merzke Reactor Projects Branch 7 |
To: | Burton C Carolina Power & Light Co |
References | |
IR-09-006 | |
Download: ML093060038 (29) | |
See also: IR 05000400/2009006
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
SAM NUNN ATLANTA FEDERAL CENTER
61 FORSYTH STREET, SW, SUITE 23T85
ATLANTA, GEORGIA 30303-8931
October 30, 2009
Mr. Christopher L. Burton
Vice President
Carolina Power & Light Company
Shearon Harris Nuclear Plant
P.O. Box 165, Mail Zone 1
New Hill, NC 27562-0165
SUBJECT: SHEARON HARRIS NUCLEAR POWER PLANT - NRC PROBLEM
IDENTIFICATION AND RESOLUTION INSPECTION
REPORT 05000400/2009006
Dear Mr. Burton:
On October 2, 2009, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection
at your Shearon Harris reactor facility. The enclosed report documents the inspection findings,
which were discussed on October 2, 2009, and October 26, 2009, with you 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, compliance with the Commissions rules and
regulations, and with the conditions of your operating license. Within these areas, the
inspection involved examination of selected procedures and representative records,
observations of plant equipment and activities, and interviews with personnel.
On the basis of the samples selected for review, the team concluded that in general, problems
were properly identified, evaluated, and resolved within the problem identification and resolution
program. However, during the inspection, some examples of minor issues were identified in the
areas of identification of issues, prioritization and evaluation of issues, and effectiveness of
corrective actions. This report documents two NRC identified findings that were evaluated
under the significance determination process as having very low safety significance (Green).
These issues were determined to involve violations of NRC requirements. However, because of
their very low safety significance and because they were entered into your corrective action
program, the NRC is treating these findings as non-cited violations consistent with
Section VI.A.1 of the NRC Enforcement Policy. If you wish to contest these non-cited 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 20555-001; with copies to the Regional Administrator Region II; the Director,
Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC
20555-0001; and the NRC Senior Resident Inspector at the Shearon Harris Nuclear Plant.
CP&L 2
In addition, if you disagree with the characterization of 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 II, and the NRC Resident Inspector at the
Shearon Harris Power Plant. The information you provide will be considered in accordance with
Inspection Manual Chapter 0305.
In accordance with 10 CFR 2.390 of the NRCs "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 the
NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Daniel Merzke, Acting Chief
Reactor Projects Branch 7
Division of Reactor Projects
Docket Nos. 50-400
License Nos. DPR-63
Enclosure: Inspection Report 05000400/2009006
w/Attachment: Supplemental Information
cc w/encl. (See page 3)
CP&L 2
In addition, if you disagree with the characterization of 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 II, and the NRC Resident Inspector at the
Shearon Harris Power Plant. The information you provide will be considered in accordance with
Inspection Manual Chapter 0305.
In accordance with 10 CFR 2.390 of the NRCs "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 the
NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Daniel Merzke, Acting Chief
Reactor Projects Branch 7
Division of Reactor Projects
Docket Nos. 50-400
License Nos. DPR-63
Enclosure: Inspection Report 05000400/2009006
w/Attachment: Supplemental Information
cc w/encl. (See page 3)
SUNSI Rev Compl. ; Yes No ADAMS ; Yes No Reviewer Initials
Publicly Avail ; Yes No Sensitive Yes ; No Sens. Type Initials
RIV:DRP RII:DRP RII:DRP RII:DRS RII:DRP
MCatts PLessard PNiebaum RTaylor EStamm
MPS4 by email PBL1 by email PKN by email RCT1 by email EJS2
10/29/09 10/29/09 10/29/09 10/29/09 10/30/09
RII:DRP RII:DRP
DMerzke RMusser
DXM2 RAM
10/30/09 10/30/09
OFFICIAL RECORD COPY DOCUMENT NAME: S:\DRP\RPB7\PI&R\PI&R\InspectionReports\Harris PIR Inspection
Report 2009006 rev 7.doc T=Telephone E=E-mail F=Fax
CP&L 3
cc w/encl:
Brian C. McCabe Chairman
Manager, Nuclear Regulatory Affairs North Carolina Utilities Commission
Progress Energy Carolinas, Inc. Electronic Mail Distribution
Electronic Mail Distribution
Beverly O. Hall
R. J. Duncan, II Chief, Radiation Protection Section
Vice President Department of Environmental Health
Nuclear Operations N.C. Department of Environmental
Carolina Power & Light Company Commerce & Natural Resources
Electronic Mail Distribution Electronic Mail Distribution
Greg Kilpatrick Public Service Commission
Training Manager State of South Carolina
Shearon Harris Nuclear Power Plant P.O. Box 11649
Progress Energy Carolinas, Inc. Columbia, SC 29211
Electronic Mail Distribution
Robert P. Gruber
John Warner Executive Director
Manager Public Staff - NCUC
Support Services 4326 Mail Service Center
Progress Energy Carolinas, Inc. Raleigh, NC 27699-4326
Electronic Mail Distribution
Herb Council
David H. Corlett Chair
Supervisor Board of County Commissioners of Wake
Licensing/Regulatory Programs County
Progress Energy P.O. Box 550
Electronic Mail Distribution Raleigh, NC 27602
David T. Conley Tommy Emerson
Associate General Counsel Chair
Legal Dept. Board of County Commissioners of
Progress Energy Service Company, LLC Chatham County
Electronic Mail Distribution 186 Emerson Road
Siler City, NC 27344
Christos Kamilaris
Director Kelvin Henderson
Fleet Support Services Plant General Manager
Carolina Power & Light Company Carolina Power and Light Company
Electronic Mail Distribution Shearon Harris Nuclear Power Plant
Electronic Mail Distribution
John H. O'Neill, Jr.
Shaw, Pittman, Potts & Trowbridge cc w/encl. (continued page 4)
2300 N. Street, NW
Washington, DC 20037-1128
CP&L 4
cc w/encl. (continued)
Senior Resident Inspector
Carolina Power and Light Company
Shearon Harris Nuclear Power Plant
U.S. NRC
5421 Shearon Harris Rd
New Hill, NC 27562-9998
CP&L 5
Letter to Christopher L. Burton from Daniel Merzke dated October 30, 2009.
SUBJECT: SHEARON HARRIS NUCLEAR POWER PLANT - NRC PROBLEM
IDENTIFICATION AND RESOLUTION INSPECTION REPORT
Distribution w/encl:
C. Evans, RII EICS
L. Slack, RII EICS
OE Mail
RIDSNRRDIRS
PUBLIC
RidsNrrPMShearonHarris Resource
U.S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos.: 50-400
License Nos.: DPR-63
Report No: 05000400/2009006
Licensee: Carolina Power and Light Company (CP&L)
Facility: Shearon Harris Nuclear Power Plant, Unit 1
Location: 5413 Shearon Harris Road
New Hill, NC 27562
Dates: September 14 - 18, 2009
September 28 - October 2, 2009
Inspectors: M. Catts, Resident Inspector, Palo Verde, Team Leader
P. Lessard, Resident Inspector, Harris
P. Niebaum, Resident Inspector, Hatch
R. Taylor, Senior Project Inspector
E. Stamm, Project Engineer
Approved by: Daniel Merzke, Acting Chief
Reactor Projects Branch 7
Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR 05000400/2009006; 09/14/2009 - 10/02/2009; Shearon Harris Nuclear Power
Plant, Unit 1; biennial inspection of the identification and resolution of problems.
The inspection was conducted by a senior project inspector, three resident inspectors, and a
project engineer. Two Green findings of very low safety significance were identified during the
inspection. The significance of most findings is indicated by their color (Green, White, Yellow,
or Red) using Inspection Manual Chapter 0609, "Significance Determination Process." The
cross-cutting aspects were determined using Inspection Manual Chapter 0305, "Operating
Reactor Assessment Program." Findings for which the significance determination process does
not apply may be Green or be assigned a severity level after NRC management's review. The
NRCs 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 inspection team concluded that, in general, problems were adequately identified, prioritized,
and evaluated; and effective corrective actions were implemented. Site management was
actively involved in the corrective action program and focused appropriate attention on
significant plant issues. The team found that employees were encouraged by management to
initiate corrective action documents to address plant issues.
The licensee generally had an adequate threshold for identifying and correcting problems, as
evidenced by the relatively few deficiencies identified by the NRC that had not been previously
identified by the licensee during the review period. Action requests normally provided complete
and accurate characterization of the problem. However, the team identified a minor violation
and seven minor issues during plant walkdowns and document reviews where problems were
not identified and entered into the corrective action program by the licensee.
Generally, prioritization and evaluation of issues were adequate, consistent with the licensees
corrective action program guidance. Formal root cause evaluations for significant problems
were adequate, and corrective actions specified for problems addressed the cause of the
problems. The age and extensions for completing evaluations were closely monitored by plant
management, both for high priority nuclear condition reports, as well as for adverse conditions
of lower priority. Also, the technical adequacy and depth of evaluations (e.g., root cause
investigations) were typically adequate. However, the team identified one unresolved item and
two minor issues associated with prioritization and evaluation of issues.
Corrective actions were generally timely, commensurate with the safety significance of the
issues, and effective, in that conditions adverse to quality were corrected in accordance with the
licensee CAP procedures. For the significant conditions adverse to quality that were reviewed,
generally the corrective actions directly addressed the cause and effectively prevented
recurrence, as evidenced by a review of performance indicators, nuclear condition reports, and
discussions with licensee staff that demonstrated that the significant conditions adverse to
quality had not recurred. Effectiveness reviews for corrective actions to prevent recurrence
were scheduled consistent with licensee procedures. However, during the review of nuclear
Enclosure
3
condition reports, the team identified two violations of NRC requirements and an additional
minor issue regarding adequacy and timeliness of corrective actions.
The operating experience program was effective in screening operating experience for
applicability to the plant, entering items determined to be applicable into the corrective action
program, and taking adequate corrective actions to address the issues. External and internal
operating experience were adequately utilized and considered as part of formal root cause
evaluations for supporting the development of lessons learned and corrective actions.
The licensees audits and self-assessments were critical and effective in identifying issues and
entering them into the corrective action program. These audits and assessments identified
issues similar to those identified by the NRC with respect to the effectiveness of the corrective
action program.
Based on general discussions with licensee employees during the inspection, targeted
interviews with plant personnel, and reviews of selected employee concerns records, the team
determined that personnel at the site felt free to raise safety concerns to management and use
the corrective action program as well as the employee concerns program to resolve those
concerns.
A. NRC Identified Findings
Cornerstone: Barrier Integrity
- Green. The team identified a non-cited violation of 10 CFR Part 50, Appendix B,
Criterion XVI, "Corrective Action," for the licensees failure to identify the cause
and take corrective actions to preclude repetition of a significant condition
adverse to quality for both containment spray additive system eductors being
outside of the technical specification flow band. Specifically, between July 2009
and the present, the violation occurred when Eductor A was found three times
and Eductor B was found once outside of the Technical Specification 3.6.2.2 flow
band. This issue was previously identified as a significant condition adverse to
quality in January 2008, but the corrective actions taken failed to preclude
repetition. The licensee entered this issue into the corrective action program as
nuclear condition report 356873. The licensee took immediate corrective actions
to throttle the eductor flow to within the band, and is developing corrective
actions to preclude repetition.
The finding is more than minor because it is associated with the design control
attribute of the Barrier Integrity Cornerstone and affects the cornerstone objective
of providing reasonable assurance that physical design barriers, such as the
iodine scrubbing capability of the containment spray additive system eductors,
will protect the public from radionuclide releases caused by accidents or events.
Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and
Characterization of Findings," the finding was determined to have a very low
safety significance because it did not represent a degradation of the radiological
barrier function provided for the control room, auxiliary building, or spent fuel
pool; the finding did not represent a degradation of the barrier function of the
Enclosure
4
control room against smoke or a toxic atmosphere; the finding did not represent
an actual open pathway in the physical integrity of reactor containment; and the
finding did not involve an actual reduction in function of the hydrogen igniters in
the reactor containment. The finding had a cross-cutting aspect in the area of
problem identification and resolution associated with the corrective action
program because the licensee did not thoroughly evaluate problems such that
the resolutions address causes and extent of conditions, as necessary, and for
significant problems, conduct effectiveness reviews of corrective actions to
ensure that the problems are resolved (P.1(c)) (Section 4OA2.a(3)(i)).
- Green. The team identified a non-cited violation of 10 CFR Part 50, Appendix B,
Criterion XVI, "Corrective Action," for the licensees failure to correct a condition
adverse to quality in a timely manner. Specifically, between May 27, 1997 and
September 29, 2007, Main Steam Isolation Valve 82 close stroke time exhibited
a condition adverse to quality for a trend degrading towards the technical
specification limit, without sufficient corrective actions to prevent failure. This
resulted in Main Steam Isolation Valve 82 exceeding the five-second stroke time
limit required in Technical Specification 3.7.1.5. The licensee entered this issue
into the corrective action program as nuclear condition report 358464.
This finding is more than minor because it is associated with the containment
barrier performance attribute of the Barrier Integrity Cornerstone and affects the
cornerstone objective of providing reasonable assurance that physical design
barriers, such as the main steam isolation valve radiological release barrier
required for a steam generator tube rupture, protect the public from radionuclide
releases caused by accidents or events. Using Manual Chapter 0609.04, "Phase
1 - Initial Screening and Characterization of Findings," the finding was
determined to have a very low safety significance because it did not represent a
degradation of the radiological barrier function provided for the control room,
auxiliary building, or spent fuel pool; the finding did not represent a degradation
of the barrier function of the control room against smoke or a toxic atmosphere;
the finding did not represent an actual open pathway in the physical integrity of
reactor containment; and the finding did not involve an actual reduction in
function of the hydrogen igniters in the reactor containment. This finding had a
cross-cutting aspect in the area of human performance associated with decision-
making because the licensee did not use conservative assumptions so that
safety-significant decisions were verified to validate underlying assumptions and
identify unintended consequences (H.1.(b)) (Section 4OA2.a(3)(ii)).
B. Licensee Identified Violations
None
Enclosure
REPORT DETAILS
4. OTHER ACTIVITIES
4OA2 Problem Identification and Resolution
a. Assessment of the Corrective Action Program
(1) Inspection Scope
The inspectors reviewed the licensees corrective action program (CAP) procedures
which described the administrative process for initiating and resolving problems primarily
through the use of action requests (ARs), which were then processed into the CAP as
nuclear condition reports (NCRs). The team selected and reviewed a sample of NCRs
that had been issued between August 2007 and August 2009. This period of time was
purposefully chosen to follow the last Biennial Problem Identification and Resolution
(PI&R) inspection conducted in August 2007. This review was performed to verify that
problems were being properly identified, appropriately characterized, and entered into
the CAP for resolution. Where possible, the team independently verified that the
corrective actions were implemented as intended.
Within the time frame described above, the team selected NCRs from principally four
specific areas of interest. The first inspection area consisted of a detailed review of
selected NCRs associated with four risk-significant systems: emergency AC power (non-
emergency diesel generator (EDG)), essential services chilled water, containment
isolation Target Rock valves, and low head safety injection (LHSI) / residual heat
removal (RHR) system. The team conducted plant walkdowns of equipment associated
with the selected systems and other plant areas to assess the material condition and to
look for any deficiencies that had not been previously entered into the CAP. The team
reviewed NCRs, maintenance history, completed work orders (WOs) for the systems,
and reviewed associated system health reports. These reviews were performed to verify
that problems were being properly identified, appropriately characterized, and entered
into the CAP for resolution. Items reviewed generally covered a two-year period of time;
however, in accordance with the inspection procedure, the team performed a five-year
review of age-dependent issues for containment isolation Target Rock valves and
LHSI/RHR.
The second inspection area consisted of a detailed review of a representative number of
NCRs that were assigned to the major plant departments, including operations,
maintenance, engineering, health physics, chemistry, emergency preparedness, and
security. This selection was performed to ensure that samples were reviewed across all
cornerstones of safety identified in the NRCs Reactor Oversight Process (ROP). These
NCRs were reviewed to assess each departments threshold for identifying and
documenting plant problems, thoroughness of evaluations, and adequacy of corrective
actions. The team also attended meetings where NCRs were screened for significance
Enclosure
6
to determine whether the licensee was identifying, accurately characterizing, and
entering problems into the CAP at an appropriate threshold.
For the third inspection area, the team selected a sample of NRC issued non-cited
violations and findings, licensee identified violations, and Licensee Event Reports
(LERs), to verify the effectiveness of the licensees CAP implementation regarding NRC
inspection findings and reportable events issued since the previous 2007 PI&R
inspection.
The fourth inspection area covered the review of NCRs associated with selected issues
of interest, specifically maintenance rule functional failures, non-conforming/degraded
conditions, and radiation monitors performance issues. The team reviewed the NCRs to
verify that problems were identified, evaluated, and resolved in accordance with the
licensees procedures and applicable NRC Regulations.
Among the four areas mentioned above, the team conducted a detailed review of
selected root-cause and apparent-cause evaluations of the problems identified. The
team reviewed these evaluations against the descriptions of the problem described in
the NCRs and the guidance in licensee Procedure CAP-NGGC-0205, "Significant
Adverse Condition Investigations and Adverse Condition Investigations-Increased
Rigor." The team assessed if the licensee had adequately determined the cause(s) of
identified problems, and had adequately addressed operability, reportability, common
cause, generic concerns, extent-of-condition, and extent-of-cause. The review also
assessed if the licensee had appropriately identified and prioritized corrective actions to
prevent recurrence.
Additionally, the team performed control room walkdowns to assess the main control
room (MCR) deficiency list and to ascertain if deficiencies were entered into the CAP.
Operator workarounds and operator burden screenings were reviewed, and the team
verified compensatory measures for deficient equipment which were being implemented
in the field.
Finally, the team reviewed site trend reports, to determine if the licensee effectively
trended identified issues and initiated appropriate corrective actions when adverse
trends were identified. The team attended various plant meetings to observe
management oversight and implementing functions of the corrective action process.
These included Management Review of NCRs meetings and Unit Evaluators meetings.
Documents reviewed are listed in the Attachment.
(2) Assessment
Identification of Issues
The team determined that the licensee generally had an adequate threshold for
identifying and correcting problems as evidenced by: the relatively few deficiencies
identified by the NRC that had not been previously identified by the licensee during the
review period; the type of problems identified and corrected; the review of licensee
Enclosure
7
requirements for initiating corrective action documents as described in licensee
Procedure CAP-NGGC-0200, "Corrective Action;" the management expectation that
employees were encouraged to initiate NCRs or work orders; a review of system health
reports; and the teams observations during plant walkdowns. However, the team
identified a minor violation and seven minor issues during plant walkdowns and
document reviews where problems were not identified and entered into the CAP by the
licensee. Trending was generally effective in monitoring and identifying plant issues;
however, the team determined that not enough time had passed to assess trends or for
the licensee to develop goals and thresholds for the newly developed performance
indicators, such as corrective maintenance backlog or preventative maintenance
deferred. Site management was actively involved in the CAP and focused appropriate
attention on significant plant issues.
The team identified the following minor violation:
- 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," states, in part, that all
testing required to demonstrate that structures, systems, and components will
perform satisfactorily in service is identified and performed in accordance with
written test procedures. It further states that test results shall be documented and
evaluated to assure that test requirements have been satisfied. Contrary to the
above, on September 30, 2009, the team identified data recorded per
Procedure MST-I0412, "Waste Processing Building (WPB) Stack 5 Flow Rate
Monitor and Isokinetic Sampling System Calibration dated August 20, 2009," was
outside the allowable range and was not discovered prior to returning the WPB Vent
Stack 5 Flow Rate Monitor and the associated Wide Range Gas Monitor (WRGM) to
service. Upon discovery, the licensee declared the WRGM inoperable and initiated
appropriate compensatory actions pending a subsequent performance of calibration
Procedure MST-I0412. This failure to comply with 10 CFR Part 50, Appendix B,
Criterion XI, "Test Control," constitutes a violation of minor significance that is not
subject to enforcement action in accordance with the NRC's Enforcement Policy.
This issue is similar to NRCs Inspection Manual Chapter 0612, Appendix E,
Example 1(a), in that the data was incorrectly recorded during the procedure and
there was reasonable assurance that the Flow Stack Monitor and the associated
WRGM remained operable as evidenced by a successful retest per licensee
Procedure MST-I0412. The licensee entered this issue into the CAP as
NCR 358187.
The team identified the following minor issues:
- The team identified a potential adverse trend in maintenance induced voiding of
safety-related systems. Specifically, voids had been introduced during maintenance
on an emergency service water (ESW) pump, a normal service water pump, a
containment spray pump, and an auxiliary feedwater pump. No operability issues
exist for these pumps. The licensee entered this issue into the CAP as NCR
356943.
- Nuclear Condition Report 357122 was written to address refrigerant/oil leakage on
Essential Services Chiller B. Per Procedure CAP-NGGC-0200, this NCR should
Enclosure
8
have been routed to the MCR so the licensee could appropriately explore any impact
upon operability. The licensee identified that the NCR had not been properly routed
to the MCR and took corrective action. However, the licensee failed to identify that
the NCR not being properly routed to the MCR was an adverse condition. Following
discussions with the inspection team, the licensee concluded that not routing the
NCR to the MCR was an adverse condition and entered the issue into the CAP as
NCR 357595.
- Emergency Diesel Generator A Frequency Transducer failed on
September 11, 2009; however, NCR 247241 was not written until nine days after the
failure. Procedure CAP-NGGC-0200 requires an NCR to be written promptly. There
was no impact to having this NCR written late. The licensee entered this issue into
the CAP as NCR 358348.
- The team reviewed the MCR logs for radiation monitor failures and discovered
Channel 2 of Radiation Monitor RM-3567ASA was declared inoperable on
June 8, 2009. During troubleshooting efforts, the licensee discovered that the
Channel 2 detector had failed. The team questioned the licensee and discovered an
NCR was not initiated to document this event. Not entering this issue into CAP had
no effect on plant equipment. The licensee entered this issue into the CAP as NCR
358412.
- During a walkdown of the RHR Trains A and B with the licensee, the inspector
identified multiple deficiencies which required entry into the CAP. The licensee
initiated NCR 355964 for obsolete testing devices remaining on motor operated valve
actuators. The licensee initiated NCR 355989 for both RHR pump vibration
monitoring cables not enclosed in flexible conduit as per design. The licensee
entered two other conditions into the CAP via work requests (WR): WR 399084 for
boric acid staining below 1RH-30 (RHR A Heat Exchanger Discharge Valve) and WR 399087 for boric acid on 1SI-359 (LHSI Supply Isolation Valve). Lastly, the licensee
initiated WR 399078 for a minor grease leak on 1SI-341 (RHR B Shutdown Cooling
Isolation Valve). The team determined that none of these issues impacted
operability of the RHR system.
- The MCR annunciator inverter power transfer setpoints were erroneously set to
104 Vdc/Vac during replacement in July 2008. This value was below the plant
drawing and vendor recommended setpoint of 120 +/- 10% Vdc/Vac. The licensee
entered this issue into the CAP as NCR 355911, determined there was no current
impact, and initiated a compensatory measure to log inverter voltage once each shift
to assure that the setpoint deficiency had no impact on the functionality of the MCR
- A safety system outage on ESW Train A, which caused a quantitative yellow risk
condition was extended and scheduled to overlap a qualitative yellow risk condition.
After this condition was identified, the licensee delayed the qualitative yellow risk
condition to prevent overlapping yellow risk conditions. The licensees
Procedure WCM-001, "On-Line Maintenance Risk Management," offered no
Enclosure
9
guidance to consider the combined effect of quantitative and qualitative risk
conditions. The licensee entered this issue into the CAP as NCR 356048.
Prioritization and Evaluation of Issues
Based on the review of audits conducted by the licensee and the assessment conducted
by the inspection team during the onsite period, the team concluded that problems were
generally prioritized and evaluated in accordance with the licensees CAP procedures as
described in the NCR Processing Guidelines in Procedure CAP-NGGC-0200. Each
NCR written was assigned a priority level at the NCR review meetings. Management
reviews of NCRs were thorough and adequate consideration was given to system or
component operability and associated plant risk.
The team determined that the station had conducted root cause and apparent cause
analyses in compliance with the licensees CAP procedures, and assigned cause
determinations were appropriate considering the significance of the issues being
evaluated. A variety of causal-analysis techniques were used depending on the type
and complexity of the issue consistent with licensee Procedure CAP-NGGC-0205.
The team determined that generally, the licensee had performed evaluations that were
technically accurate and of sufficient depth. The team further determined that
operability, reportability, and degraded or non-conforming condition determinations had
been completed consistent with the guidance contained in Procedures CAP-NGGC-0200
and OPS-NGGC-1305, "Operability Determinations." However, the team identified one
unresolved item (URI) which is documented in Section 4OA2.a(3)(iii) of this report, and
two minor issues in this assessment area during the review of NCRs:
- Emergency Diesel Generator A Frequency Transducer failed on
September 11, 2009; however, the licensee determined a reportability review was
not required for the failed component as documented in NCR 247241.
Procedure CAP-NGGC-0200 requires NCRs be reviewed for reportability. The
licensee performed a preliminary review and determined that the frequency
transducer failed in a conservative direction. The licensee entered this issue into the
CAP as NCR 357786.
- Nuclear Condition Report 263267 investigated the degraded grid time delay relays
for the safety-related 6.9 kilovolt (kV) Busses 1A-SA and 1B-SB that failed their
as-found TS surveillance test during refueling outage (RFO) 14. The team
questioned the licensee on their selected cause for the relay failures and determined
that the defective relays were not quarantined or evaluated, following their
replacement, in an effort to validate the selected cause. The licensee entered this
issue into the CAP as NCR 358290 to improve the quarantine process for defective
parts. The team concluded that the selected cause was adequate based on
available information and that corrective action to replace the failed relays with a
different type of relay was adequate.
Enclosure
10
Effectiveness of Corrective Actions
Based on a review of corrective action documents, interviews with licensee staff, and
verification of completed corrective actions, the team determined that overall, corrective
actions were timely, commensurate with the safety significance of the issues, and
effective, in that conditions adverse to quality were corrected in accordance with the
licensee CAP procedures. For the significant conditions adverse to quality reviewed,
generally the corrective actions directly addressed the cause and effectively prevented
recurrence, as evidenced by a review of performance indicators, NCRs, and discussions
with licensee staff that demonstrated that the significant conditions adverse to quality
had not recurred. Effectiveness reviews for corrective actions to preclude recurrence
(CAPRs) were scheduled consistent with licensee procedures. However, during the
review of NCRs, the team identified two violations of NRC requirements and an
additional minor issue regarding adequacy and timeliness of corrective actions.
The team identified the following two violations:
- Between July 2009 and the present, Containment Spray Additive System Eductor A
was found three times and Eductor B was found once outside of the TS 3.6.2.2 flow
band. This issue was previously identified as a significant condition adverse to
quality in January 2008, but the corrective actions taken failed to preclude
recurrence. The team identified one finding for the failure to identify the cause and
take CAPR of a significant condition adverse to quality for both containment spray
additive system eductors being outside of the TS flow band as documented in
Section 4OA2.a(3)(i). The licensee entered this issue into the CAP as NCR 356873.
- Between May 27, 1997 and September 29, 2007, Main Steam Isolation Valve MS-82
close stroke time exhibited a degrading trend towards the TS limit without sufficient
corrective actions to prevent failure. This resulted in MS-82 exceeding the five-
second stroke time limit required in TS 3.7.1.5. The team identified one finding for
failure to correct a condition adverse to quality in a timely manner as documented in
Section 4OA2.a(3)(ii). The licensee entered this issue into the CAP as NCR 358464.
The team identified the following minor issue:
- Nuclear Condition Report 290961 evaluated the failure of the main condenser
expansion joint that caused a loss of vacuum and resulted in a manual trip of the
unit. This issue was discussed in more detail in LER 2008-002-00. The team
determined that while the corrective actions were generally adequate, the expansion
joint inspection instructions do not contain specific acceptance criteria. Specific
acceptance criteria for inspecting for dry rot, cracking, splitting or other signs of
degradation is necessary to ensure an objective review to determine if results are
satisfactory. The team determined that the potential still exists for degradation not
being properly identified. The licensee entered this issue into the CAP as NCR
358345.
Enclosure
11
(3) Findings
(i) Failure to Preclude Repetition of a Significant Condition Adverse to Quality for Both
Containment Spray Additive System Eductors Being Outside of the Technical
Specification Flow Band
Introduction. The team identified a Green non-cited violation of 10 CFR Part 50,
Appendix B, Criterion XVI, "Corrective Action," for the licensees failure to identify the
cause and take CAPR of a significant condition adverse to quality for both containment
spray additive system eductors being outside of the TS flow band, which resulted in
Eductor A found three times and Eductor B found once outside of the TS 3.6.2.2 flow
band between July 2009 and the present.
Description. Between November 2007 and May 2008, the containment spray additive
system eductors were found outside of the TS 3.6.2.2 flow band seven times. In
January 2008, the licensee determined that this was a significant condition adverse to
quality and performed a root cause investigation. During the course of their
investigation, the licensee identified two root causes: entrapped air in the system and
inadequate system design. As CAPRs, the licensee established a procedure to identify
air voids in the system, revised the operations procedure to prevent the eductors from
being operated with the suction line isolated, and installed more stable throttle valves in
the suction line. The licensee reported the condition to the NRC in May 2008 as
LER 2008-01-00. This LER was closed as a Licensee Identified Violation (LIV) in
Inspection Report 05000400/2008004.
The purpose of the eductor is to introduce sodium hydroxide (NaOH) into the
containment spray (CT) system flow during a loss of coolant accident. If there is too little
eductor flow, not enough NaOH would be present and the iodine scrubbing capability of
the CT system would be reduced. If too much NaOH is present, CT flow pH could rise
high enough to increase degradation of aluminum in containment. This could result in
increased debris accumulating on the emergency core cooling system recirculation
sump screens and reducing performance of the emergency core cooling system. During
their previous investigation, the licensee determined that they had experienced eductor
flows both above and below the TS flow band.
The team reviewed the licensees implementation of the CAPRs, and determined the
CAPRs were ineffective at precluding repetition of a significant condition adverse to
quality since the eductor flows were discovered outside of the TS band between
July 2009 and the present. On three occasions flow was below the TS band, and on one
occasion flow was above the TS band. The licensee took immediate corrective actions
to adjust flow back into the TS band. Additionally, the licensee developed a
compensatory measure to dispatch a dedicated operator to adjust flow as necessary in
the case of CT initiation. The licensee initiated NCR 356873, reopened the root cause
investigation, is reevaluating the cause determination that was performed in 2008, and is
developing additional CAPRs to address the root cause.
Analysis. The performance deficiency associated with this finding involved the
licensees failure to identify the cause and take CAPR of a significant condition adverse
Enclosure
12
to quality, resulting in both containment spray additive system eductors being outside of
the TS 3.6.2.2 flow band. The finding is more than minor because it is associated with
the design control attribute of the Barrier Integrity Cornerstone and affects the
cornerstone objective of providing reasonable assurance that physical design barriers,
such as the iodine scrubbing capability of the containment spray additive system
eductors, will protect the public from radionuclide releases caused by accidents or
events. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and
Characterization of Findings," the finding was determined to have a very low safety
significance because it did not represent a degradation of the radiological barrier
function provided for the control room, auxiliary building, or spent fuel pool; the finding
did not represent a degradation of the barrier function of the control room against smoke
or a toxic atmosphere; the finding did not represent an actual open pathway in the
physical integrity of reactor containment; and the finding did not involve an actual
reduction in function of the hydrogen igniters in the reactor containment. The finding has
a cross-cutting aspect in the area of problem identification and resolution associated with
the corrective action program because the licensee did not thoroughly evaluate
problems such that the resolutions address causes and extent of conditions, as
necessary, and for significant problems, conduct effectiveness reviews of corrective
actions to ensure that the problems are resolved (P.1(c)).
Enforcement. Title 10 of the Code of Federal Regulations, Part 50, Appendix B,
Criterion XVI, "Corrective Action," requires, in part, that in the case of a significant
condition adverse to quality, the measures taken shall assure that the cause of the
condition is determined and corrective action should preclude repetition. Contrary to this
requirement, the licensee failed to identify the cause and take CAPR of both
containment spray additive system eductors being outside of the TS flow band.
Specifically, between July 2009 and the present, the violation occurred when Eductor A
was found three times and Eductor B was found once outside of the TS 3.6.2.2 flow
band.
The licensee took immediate corrective action to throttle eductor flow to within the TS
band, and is developing CAPRs. Because the finding is of very low safety significance
and has been entered into the licensees CAP as NCR 356873, this violation is being
treated as an NCV consistent with Section VI.A.1 of the Enforcement Policy:
NCV 05000400/ 2009006-01, "Failure to Preclude Repetition of a Significant Condition
Adverse to Quality for Both Containment Spray Additive System Eductors Being Outside
of the Technical Specification Flow Band."
(ii) Failure to Correct a Condition Adverse to Quality Involving a Main Steam Isolation Valve
Degrading Trend Before Valve Failure
Introduction. The team identified a Green non-cited violation of 10 CFR Part 50,
Appendix B, Criterion XVI, "Corrective Action," for the licensees failure to correct a
condition adverse to quality in a timely manner, which resulted in MS-82 exceeding the
TS stroke time limit.
Description. On September 29, 2007, Valve MS-82 failed surveillance test
Procedure OST-1046, "Main Steam Isolation Valve Operability Test Quarterly Interval
Enclosure
13
Mode 3 to 5," due to exceeding the close stroke time limit of five seconds. Technical
Specification Surveillance Requirement 4.7.1.5, "Main Steam Line Isolation Valves,"
requires this valve to stroke close within five seconds. The main steam isolation valves
are required to close to act as a barrier to a radiological release during a steam
generator tube rupture or to mitigate a main steam line break. The licensee declared
Valve MS-82 inoperable, wrote NCR 248429, and performed WO 1120864 to repair the
valve and decrease the stroke time.
The licensee had been trending the close stroke time of Valve MS-82 since
December 29, 1986. The close stroke time trend started to degrade around
May 27, 1997. In May 2004, the valve was labeled low margin due to the valve stroking
close at 4.74 seconds, which was approaching the five-second limit. Between May 2004
and RFO 13 in April 2006, the valve stroke time continued to increase so that at the start
of RFO 13 the valve stroked close at 4.96 seconds. The licensee replaced the actuator
of the valve; however, the as-left valve stroke time at the end of RFO 13 was still near
the TS limit at 4.92 seconds.
The licensee developed contingency WO 1120864 for RFO 14, to gain stroke time
margin by adjusting the air operated valve hydraulic system flow control valve. During
RFO 14, on September 29, 2007, Valve MS-82 failed the stroke time close test by
stroking at 5.17 seconds. The licensee implemented contingency WO 1120864.
The team reviewed NCR 248429 and the close stroke time trend for Valve MS-82. The
team questioned why the degrading trend since 1997 had not been identified, and an
NCR had not been written to correct the condition. The team determined that unlike the
other valves in the in-service testing program, no process or procedure existed to
identify a degrading trend on a main steam isolation valve, write a NCR, and correct the
condition before valve failure. The team determined this issue was indicative of current
plant performance since no process or procedure currently exists.
The team questioned that with the degrading trend nearing the close stroke time limit,
why effective maintenance was not performed in RFO 13 to ensure the valve would not
exceed the TS close stroke time before RFO 14. The team reviewed the surveillance
test performed on April 8, 2006, and noted that the licensee was still in Mode 5 where
maintenance could have been performed on the valve. However, the team noted that
the surveillance test results were not reviewed until April 11, 2006, when the plant was in
Mode 3, when maintenance could not be performed on the valve. The team also
reviewed NCR 248429 that stated "It consistently has been a conscious decision not to
adjust these valves to gain stroke time margin because of the ensuing post maintenance
test required." This NCR also stated that the decision not to perform maintenance was
deemed to be an acceptable risk. Not performing effective maintenance on the
degrading stroke time close trend for Valve MS-82 led to the failure of this valve in
RFO 14. The licensee wrote NCR 358464 to address why corrective actions were not
taken before Valve MS-82 failed.
Analysis. The performance deficiency associated with this finding involved the
licensees failure to correct a condition adverse to quality in a timely manner, which
resulted in Valve MS-82 exceeding the TS stroke time limit. This finding is more than
Enclosure
14
minor because it is associated with the containment barrier performance attribute of the
Barrier Integrity Cornerstone and affects the cornerstone objective of providing
reasonable assurance that physical design barriers, such as the main steam isolation
valve radiological release barrier required for a steam generator tube rupture, protect
the public from radionuclide releases caused by accidents or events. Using Manual
Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the
finding was determined to have a very low safety significance because it did not
represent a degradation of the radiological barrier function provided for the control room,
auxiliary building, or spent fuel pool; the finding did not represent a degradation of the
barrier function of the control room against smoke or a toxic atmosphere; the finding did
not represent an actual open pathway in the physical integrity of reactor containment;
and the finding did not involve an actual reduction in function of the hydrogen igniters in
the reactor containment. This finding has a cross-cutting aspect in the area of human
performance associated with decision-making because the licensee did not use
conservative assumptions so that safety-significant decisions were verified to validate
underlying assumptions and identify unintended consequences (H.1.(b)).
Enforcement. Title 10 of the Code of Federal Regulations, Part 50, Appendix B,
Criterion XVI, "Corrective Action," requires, in part, that measures shall be established
to assure that conditions adverse to quality are promptly identified and corrected.
Contrary to this requirement, between May 27, 1997 and September 29, 2007, the
licensee failed to identify and correct a condition adverse to quality for a trend degrading
towards the technical specification limit, without sufficient corrective actions to prevent
failure. This resulted in Main Steam Isolation Valve 82 exceeding the five-second stroke
time limit required in Technical Specification 3.7.1.5. Because the finding is of very low
safety significance and has been entered into the licensees CAP as NCR 358464, this
violation is being treated as an NCV consistent with Section VI.A.1 of the Enforcement
Policy: NCV 05000400/2009006-02, "Failure to Correct a Condition Adverse to Quality
Involving a Main Steam Isolation Valve Degrading Trend Before Valve Failure."
(iii) Unresolved Item Associated With the Evaluation of the Failure of Emergency Service
Water Valve 271
Introduction. The inspectors identified a URI associated with the evaluation of the failure
of ESW Auxiliary Reservoir Discharge Valve 271 to open on the start of ESW Pump B.
Description. On October 19, 2007, while in Mode 5, ESW Auxiliary Reservoir Discharge
Valve 271 failed to open on the start of ESW Pump B. This valve is required to open on
the start of an ESW pump to provide a discharge path for the cooling water. Operators
immediately stopped ESW Pump B and aligned normal service water to the safety
related components in Train B. The licensee determined that the auto open controls for
Valve SW-271 had been disabled by a clearance order for unrelated work. Although
ESW Train B is not required to be operational in Mode 5, the components cooled by
ESW Train B, such as EDG B and RHR Train B, were being relied upon as protected
train equipment. Therefore, ESW Train B was necessary to ensure core decay heat
removal in the event that off-site power was not available. NRC inspectors wrote a
self-revealing NCV of TS 6.8.1, "Programs and Procedures," for an inadequate
clearance order as documented in NRC Integrated Inspection Report
Enclosure
15
05000400/2007005. The team reviewed the evaluation performed for this NCV including
the reportability review. The reportability review stated this condition was not reportable
since operators were able to open this valve manually from the control room. The team
questioned whether the operators would be able to open the valve within one minute,
which is required to ensure cooling to the EDGs during an accident. The team also
determined that when the valve is manually opened by the reactor operators from the
control room, that the valve would automatically go closed due to the inadequate
clearance. As a result of the teams questions, the licensee wrote NCR 358062 and
determined that the failure of SW-271 to open was a MRFF. This failure did not exceed
the ESW Train B maintenance rule performance criteria. The licensee determined that
this failure affected the MSPI. This condition could prevent the fulfillment of the safety
function of EDG B and RHR B that are needed to maintain the reactor in a safe
shutdown condition or to remove residual heat. The licensee wrote NCR 361821 to
address this issue. This issue is considered unresolved pending additional NRC review
of the evaluation of the failure including the reportability review, the risk assessment, and
the corrective actions: URI 05000400/2009006-03, "Unresolved Item Associated with
the Evaluation of the Failure of Emergency Service Water Valve 271."
b. Assessment of the Use of Operating Experience
(1) Inspection Scope
The team examined licensee programs for reviewing industry operating experience
(OE), reviewed licensees Procedure CAP-NGGC-0202, "Operating Experience
Program," and reviewed the licensees OE database, to assess the effectiveness of how
external and internal OE data was handled at the plant. In addition, the team selected
OE documents (e.g., NRC generic communications, 10 CFR Part 21 reports, LERs,
vendor notifications, etc.), which had been issued since August 2007, to verify whether
the licensee had appropriately evaluated each notification for applicability to the Shearon
Harris Nuclear Power Plant, and whether issues identified through these reviews were
entered into the CAP.
Documents reviewed are listed in the Attachment.
(2) Assessment
Based on interviews and a review of documentation related to the review of OE issues,
the team determined that the licensee was generally effective in screening OE for
applicability to the plant. Industry OE was evaluated at either the corporate or plant level
depending on the source and type of document. Relevant information was then
forwarded to the applicable department for further action or informational purposes.
Operating experience issues requiring action were entered into the CAP for tracking and
closure. In addition, OE was included in apparent cause and root cause evaluations in
accordance with licensee Procedure CAP-NGGC-0205.
(3) Findings
No findings of significance were identified.
Enclosure
16
c. Assessment of Self-Assessments and Audits
(1) Inspection Scope
The team reviewed audit reports and self-assessment reports, including those which
focused on problem identification and resolution, to assess the thoroughness and
self-criticism of the licensee's audits and self-assessments, and to verify that problems
identified through those activities were appropriately prioritized and entered into the CAP
for resolution in accordance with licensee Procedure CAP-NGGC-0201,
"Self-Assessment and Benchmark Programs."
(2) Assessment
The team determined that the scopes of assessments and audits were adequate.
Self-assessments were generally detailed and critical, as evidenced by findings
consistent with the teams independent review. Self-assessment findings related to
issues or weaknesses were entered into the CAP and tracked to completion based on
the NCR priority level. Corrective actions for self-assessment findings were adequate to
address the issues. Generally, the licensee performed evaluations that were technically
accurate. Site trend reports were thorough and a low threshold was established for
evaluation of potential trends; however, the team determined that not enough time had
passed to assess trends or for the licensee to develop goals and thresholds for the
newly developed performance indicators, such as corrective maintenance backlog or
preventative maintenance deferred. The team concluded that the self-assessments and
audits were an effective tool to identify adverse trends.
(3) Findings
No findings of significance were identified.
d. Assessment of Safety-Conscious Work Environment
(1) Inspection Scope
The team randomly interviewed 29 on-site workers from maintenance, security,
operations, chemistry, and engineering organizations regarding their knowledge of the
corrective action program at Shearon Harris and their willingness to write NCRs or raise
safety concerns. During technical discussions with members of the plant staff, the team
conducted interviews to develop a general perspective of the safety-conscious work
environment at the site. The interviews were also conducted to determine if any
conditions existed that would cause employees to be reluctant to raise safety concerns.
The team reviewed the licensees employee concerns program (ECP) and interviewed
the ECP coordinator. Additionally, the team reviewed the latest Safety Culture
Assessment to evaluate the thoroughness and self-criticism of the licensee's
assessment, and to verify that problems identified were appropriately prioritized and
entered into the CAP for resolution. Finally, the team reviewed a sample of completed
ECP reports to verify that concerns were being properly reviewed and identified
deficiencies were being resolved and entered into the CAP when appropriate.
Enclosure
17
(2) Assessment
Based on the interviews conducted and the NCRs reviewed, the team determined that
licensee management emphasized the need for all employees to identify and report
problems using the appropriate methods established within the administrative programs,
including the CAP and ECP. These methods were readily accessible to all employees.
Based on discussions conducted with a sample of plant employees from various
departments, the team determined that employees felt free to raise issues, and that
management encouraged employees to place issues into the CAP for resolution. The
team did not identify any reluctance on the part of the licensee staff to report safety
concerns.
(3) Findings
No findings of significance were identified.
4OA6 Meetings, Including Exit
On October 2, 2009, the team presented the inspection results to Mr. Christopher Burton
and other members of the site staff. On October 26, 2009, the team lead re-exited the
inspection results concerning the unresolved item to Mr. Dave Corlett.
The team confirmed that all proprietary information reviewed was returned to the
licensee during the inspection.
ATTACHMENT: SUPPPLEMENTAL INFORMATION
Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
B. Bernard, Superintendent, Security
C. Burton, Vice President Harris Plant
D. Corlett, Supervisor, Licensing/Regulatory Programs
J. Dills, Manager, Operations
J. Doorhy, Licensing
K. Harshaw, Manager, Outage and Scheduling
K. Henderson, Plant General Manager
J. Jankens, Supervisor, Radiation Control
G. Kilpatrick, Training Manager
P. Morales, Employee Concerns Program
L. Morgan, Supervisor, Self Evaluation Unit
S. OConnor, Manager, Engineering
M. Parker, Superintendent, Radiation Protection
B. Parks, Manager, Nuclear Oversight Section
J. Robinson, Superintendent, Environmental and Chemistry
H. Szews, CAP Coordinator
J. Warner, Manager, Support Services
NRC
J. Austin, Senior Resident Inspector
R. Musser, Chief, Reactor Projects Branch 4, Division of Reactor Projects, Region II
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
05000400/2009006-01 NCV Failure to Preclude Repetition of a Significant
Condition Adverse to Quality for Both Containment
Spray Additive System Eductors Being Outside of the
Technical Specification Flow Band (Section
4OA2.a(3)(i))05000400/2009006-02 NCV Failure to Correct a Condition Adverse to Quality
Involving a Main Steam Isolation Valve Degrading
Trend Before Valve Failure (Section 4OA2.a(3)(ii))
Opened
05000400/2009006-03 URI Unresolved Item Associated with the Evaluation of the
Failure of Emergency Service Water Valve 271
(Section 4OA2.a(3)(iii))
Closed
None
Discussed
None
Attachment
LIST OF DOCUMENTS REVIEWED
Procedures
ADM-NGGC-0113, Performance Planning and Monitoring, Revision 0
ADM-NGGC-0101, Maintenance Rule Program, Revision 20
ADM-NGGC-0104, Work Management Process, Revision 33
AP-013, Plant Nuclear Safety Committee, Revision 34
AP-930, Plant Observation Program, Revision 10
AOP-022, Loss of Service Water, Revision 29
OPS-NGGC-1305 Operability Determinations, Revision 1
CAP-NGGC-0200, Corrective Action Program, Revision 27
CAP-NGGC-0201, Self Assessment and Benchmark Programs, Revision 12
CAP-NGGC-0202, Operating Experience Program, Revision 15
CAP-NGGC-0205, Significant Adverse Condition Investigations and Adverse Condition
Investigations - Increased Rigor, Revision 9
CAP-NGGC-0206, Corrective Action Program Trending and Analysis, Revision 3
NOS-NGGC-0400, Employee Concerns Program, Revision 0
EGR-NGGC-0010, System & Component Trending Program and System Notebooks,
Revision 13
ISI-801, Inservice Testing of Valves, Revision 47
HESS Standards, Revision 5
OST-1046, Main Steam Isolation Valve Operability Test Quarterly Interval Mode 3 to 5,
Revision 12
PLP-624, Mechanical Equipment Qualification Program, Revision 18
OP-148, Essential Services Chilled Water System, Revisions 37 and 49
HPS-NGGC-0003, Radiological Posting, Labeling and Surveys, Revision 14
MST-E0045, 6.9 KV Emergency Bus 1A-SA and 1B-SB Under Voltage Relay Channel
Calibration, Revision 23
ADM-NGCC-0203, Preventative Maintenance and Surveillance Testing Administration,
Revision 13
OST-1124, Train B 6.9 KV Emergency Bus Undervoltage Trip Actuating Device Operational
Test and Contact Check Modes 1-6, Revision 25
HPS-NGGC-1000, Radiation Protection and Conduct of Operations, Revision 0
SP-013 Administrative/Support Key and Lock Control, Revision 12
AP-504 Administrative Controls for Locked and Very High Radiation Areas, Revision 29
PLP-511 Radiation Control and Protection Program, Revision 20
CRC-240 Plant Vent Stack 1 Effluent Sampling, Revision 11
HNPS-NGGC-0003, Radiological Posting, Labeling and Surveys, Revision 14
MST-E0075, 6.9 KV Emergency Buses, 1A-SA and 1B-SB Undervoltage (Loss of Voltage)
Channel Calibration, Revision 6
NGGM-IA-0038, Carolinas - Nuclear Generation Group Siren Maintenance, Revision 1
ERC-004, Environmental and Chemistry Administrative Guidelines, Revision 25
SEC-NGGC-2120, Protection of Safeguards Information, Revision 22
WCM-001, On-Line Maintenance Risk Management, Revision 20
OST-1118, Containment Spray Operability Train A Quarterly Interval Modes 1-4, Revision 33
OST-1119, Containment Spray Operability Train B Quarterly Interval Modes 1-4, Revision 35
MST-I0019, Main Steam/Feedwater Flow Loop 2 Channel Calibration, Revision 16
ADM-NGGC-0104, Work Management Process, Revision 33
MMM-002, Corrective Maintenance, Revision 17
Attachment
3
MNT-NGGC-1000, Fleet Conduct of Maintenance, Revision 0
WCM-005, Work Order Prioritization Process, Revision 8
Completed Surveillance Tests
OST-1046, Main Steam Isolation Valve Operability Test Quarterly Interval Mode 3 to 5,
Revision 12, September 29, 2007
OST-1046, Main Steam Isolation Valve Operability Test Quarterly Interval Mode 3 to 5,
Revision 12, May 11, 2006
MST-I0412, Waste Processing Building (WPB) Stack 5 Flow Rate Monitor and Isokinetic
Sampling System Calibration, August 20, 2009
Action Requests/Nuclear Condition Reports
223911 244705 245320 245633 246582 247241
248429 250575 250810 262037 263421 266234
269409 279287 279715 281217 286843 297210
300052 300163 301267 315670 318483 320236
320444 323631 329044 330455 337027 338184
340240 340325 230031 238372 238374 263439
263441 270215 282037 287726 249284 330423
301267 329438 331701 346484 282037 279704
358062 350078 251296 249347 357786 250810
279715 244705 249347 344729 266234 248429
249992 253347 257853 262001 262192 263486
265063 267065 267066 267080 267244 268566
269406 271452 275878 278486 280015 281538
285149 285222 290761 299832 306876 316594
319422 333716 196258 221803 222730 224208
228947 253347 314660 301267 300163 286843
280649 279988 277165 269409 251296 249347
266234 263921 250810 248429 247241 244705
246582 262037 245320 245633 281217 330455
279715 231046 303142 211360 246397 292892
332141 334996 246397 292892 334934 334167
334937 263267 334936 249331 316381 253376
245663 286104 288188 326920 310739 226843
267946 307600 340516 329378 352310 283579
274978 255529 330676 241895 261182 231941
328537 201481 229805 248378 226843 327372
301730 315269 171602 188528 191359 197522
207516 223563 225187 236248 243993 246188
247129 251191 252290 254402 258053 258053
261182 263759 270318 274708 279681 281080
291651 292337 305661 313305 323057 331371
349905 350640 351437 351623 351623 355964
355989 244576 248430 252234 252471 264812
302079 317205 317280 329488 329489 331169
333828 333830 336394 340319 310373 336342
336569 247193 251437 266063 278730 279326
Attachment
4
297789
Operating Experience Action Requests
306876 317361 327306 297210 329044 337027
234055 270275 291396 291403 302656 306234
Audits and Self-Assessment Items
07-16-SP-H, HNP Nuclear Safety Culture Assessment, June 6, 2007
H-SE-06-01, Harris Site Wide Self Evaluation, June 20, 2006
H-SE-08-01, Harris Nuclear Plant Self Evaluation and Human Performance Assessment,
June 16, 2008
H-OP-09-01, Assessment of Harris Operations Program, September 14, 2009
H-OM-FR-09-03, Focused Review of Return to Service Plans, January 19-23, 2009
H-MC-08-01, Harris Nuclear Material and Contact Services Assessment, February 7, 2008
H-MA-08-01, Harris Nuclear Plant Maintenance Assessment, July 2, 2008
H-TQ-07-01, Harris Nuclear Plant Training and Qualification Assessment, May 18, 2007
216880, Maintenance Procedure Backlog and Quality, August 6-10, 2009
312544, RFO-15 Post Outage Self Assessment, May 18 - June 15, 2009
314117, Harris Mid-Cycle Assessment, January 26 - February 6, 2009
264521, Closed Systems With the Source of Demineralized Water, June 2 - 5, 2008
H-ES-09-01, Harris Engineering Support Section Assessment
H-EC-08-01, HNP Environmental and Chemistry, Assessment, April 9, 2008
H-EC-06-01, HNP Environmental and Chemistry, Assessment, April 25, 2006
H-FR-07-03, Results of Environmental and Chemistry Review, January 28, 2008
H-EP-08-01, HNP Emergency Preparedness Assessment, September 26, 2008
H-EP-07-01, HNP Emergency Preparedness Assessment, October 15, 2007
H-SC-08-01, HNP Security Assessment, May 29, 2008
H-SC-07-01, HNP Security Assessment, June 14, 2007
Effectiveness Reviews
250171 226902 225952 222534 206710 201667
Work Orders
01299014 01083809 01083013 01407305 01432464 01007488
01301181 01536832 01116354 01172181 01154591 01432540
01557072 01579680 01581990 01581962 01503467 01120864
00417204 01150648 01284574 01293105 01300467 01300968
01346720 01346721 01363224 01396056 01396242 01496138
01500794 01542758 01544206 00103940 794838 1057227
1062572 1137107 1463763 1457995 1548788 769595
769599 1342247 1342249 1342251 1136753 1527115
1527116 1402107 1076326 1070000 1133326 1379777
1291028 1439053 1535610 1367060 1552520
Engineering Changes
EC66198, Evaluation of R14 UT Results of Service Water Piping, Revision 0
EC69988, Replace Isokinetic Sampling Skid, Revision 3
Attachment
5
Other Documents
Site Key Performance Indicators, January - August, 2009
Daily Management Review Meeting Agenda, September 15 and 16, 2009
Joint Steering Committee and Core Team Meeting Agenda, June 2 and 4, 2009
Key Performance Indicators for Site Human Performance, January - August, 2009
Clearance Order 153137, R14 Smoke Damper Installation, October 8, 2007
Clearance Order 108581, Replace Piston Actuator on 1MS-82, April 14, 2006
Harris Shift Narrative Log, October 8 - 19, 2007
Stroke Time Trend Data for 1SW-40, 1SW-271, and 1SW-274, October 2007
Harris Relief Request I3R-05, 2008
Drawing 2166-B-401, Service Water System B Miscellaneous Alarms, Sheet 2232
Drawing 2166-B-401, Auxiliary Transfer Panel, Sheets 822, 835, 842, 847, 846, 3297
Harris Nuclear Safety Culture Assessment, June 6, 2007
Harris Nuclear Safety Culture Debrief Notes, September 14-18, 2009
Harris Shift Narrative Log, October 14-16, 2007
Calculation CT-0063, Void Size Acceptance Criteria for Presence of Air within the Containment
Spray Additive System, Revision 0
Calculation HNP-M/Mech-1095, Limiting Void Sizes for Containment Spray Suction Piping,
Revision 0
Drawing CPL-2165, S-0550, Containment Spray System, Revision 16
NUREG-1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73, Revision 2
Main Steam Isolation Valves 80, 82, and 84 Closed Stroke Time Trends, 2001-2009
4085 - Essential Services Chilled Water System Health Report, July 28, 2009
ESCW Preventative Maintenance for 2007, September 30, 2009
3Q07 - 4Q08 Site Trend Reports, Self Evaluation Rollup and Trend Analysis
Plant Nuclear Safety Committee Action Items, July 15, 2009
Nuclear Safety Review Committee Meeting Minutes, August 21, 2007, October 29, 2007,
June 3, 2008, August 19, 2008
SD-148, System Description, Essential Services Chilled Water, Revision 15
DBD-132, Design Basis Document, Essential and Nonessential Services Chilled Water,
Revision 10
Drawing 5-S-0998, Simplified Flow Diagram, HVAC Essential Services Chilled Water,
Revision 7
CPL 2166 S-0302, Medium Voltage Relay Settings 6900V Emer. Bus 1A-SA Sheets 20, 23 and
24, Revision 9
SD-156, Plant Electrical Distribution System Description, Revision 13
System Health Report 6.9KV AC Distribution, 1st Quarter 2009, July 20, 2009
System Health Report Radiation Monitoring, 1st Quarter 2009, July 14, 2009
Calculation E2-0005.09 Degraded Grid Voltage Protection For 6.9 kV Busses 1A-SA & 1B-SB,
Revision 2
CAR-SH-N-029, Safety-Related Radiation Monitoring System Specification, Revision 6
System 5145 (Startup and Auxiliary Transformers) Maintenance Rule Scoping Document
System 5165 (6.9 KV AC Distribution) Maintenance Rule Scoping Document
STGP 208986 - Strategic Plan to replace 6.9kV air circuit breakers with vacuum breakers
Westinghouse Technical Bulletin TB-07-5, May 14, 2007
SD-118, Radiation Monitoring System Description, Revision 10
DBD-304, Radiation Monitoring System and Gross Failed Fuel Detector Design Basis
Document, Revision 9
Attachment
6
Preventative Maintenance Requests 253955, 313698
Calculation 0054-JRG, PSB-1 Loss of Offsite Power Relay Settings, Revision 3
Maintenance Rule Expert Panel meeting summary, November 15, 2007
Harris Main Condenser Trending Basis Document
Harris Nuclear Plant Emergency Preparedness Zone Siren Acoustic Study
Harris Emergency Preparedness Siren Battery Backup Power Calculations
Areva, Shearon Harris End of Cycle 15 Fuel Inspection Results
Environmental and Chemistry - Leadership Improvement Plan
Environmental and Chemistry - Self Evaluation Overview
Drawing 2165-S-0550, Simplified Flow Diagram Containment Spray System
Containment Spray System Troubleshooting Plan, September 17, 2009
Calculation CT-0027, Detail Calculation of NaOH Eductor Loop
LER 2008-003-00, Manual actuation of the Reactor Protection System During Shutdown Rod
Position Indication Surveillance testing
LER 2007-002-00, Control Rod Shutdown Bank Anomaly Causes Entry into TS 3.0.3
LER 2008-002-00, Manual Actuation of the Reactor Protection System due to Main Condenser
Exhaust Boot Failure
LER 2008-001-00, Containment Spray Additive System Eductor Test Flow Outside of TS limits
HNP Shift Narrative Log, September 17, 2009
Steam Generator Blowdown System Training Manual, Revision 5
9001-Containment Isolation Valve Health Report. July 23, 2009
EIR 20090373, Equipment Inoperable Record 1SP-217, May 19, 2009
DBD-101, Reactor Coolant Sampling, Revision 5
Operator Challenges Log, August 2009
Attachment