ML093060038

IR 05000400-09-006; 09/14/2009 - 10/02/2009; Shearon Harris Nuclear Power Plant, Unit 1; Biennial Inspection of the Identification and Resolution of Problems

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

Manager, Nuclear Regulatory Affairs

Progress Energy Carolinas, Inc.

Electronic Mail Distribution

R. J. Duncan, II

Vice President

Nuclear Operations

Carolina Power & Light Company

Electronic Mail Distribution

Greg Kilpatrick

Training Manager

Shearon Harris Nuclear Power Plant

Progress Energy Carolinas, Inc.

Electronic Mail Distribution

John Warner

Manager

Support Services

Progress Energy Carolinas, Inc.

Electronic Mail Distribution

David H. Corlett

Supervisor

Licensing/Regulatory Programs

Progress Energy

Electronic Mail Distribution

David T. Conley

Associate General Counsel

Legal Dept.

Progress Energy Service Company, LLC

Electronic Mail Distribution

Christos Kamilaris

Director

Fleet Support Services

Carolina Power & Light Company

Electronic Mail Distribution

John H. O'Neill, Jr.

Shaw, Pittman, Potts & Trowbridge

2300 N. Street, NW

Washington, DC 20037-1128

Chairman

North Carolina Utilities Commission

Electronic Mail Distribution

Beverly O. Hall

Chief, Radiation Protection Section

Department of Environmental Health

N.C. Department of Environmental

Commerce & Natural Resources

Electronic Mail Distribution

Public Service Commission

State of South Carolina

P.O. Box 11649

Columbia, SC 29211

Robert P. Gruber

Executive Director

Public Staff - NCUC

4326 Mail Service Center

Raleigh, NC 27699-4326

Herb Council

Chair

Board of County Commissioners of Wake

County

P.O. Box 550

Raleigh, NC 27602

Tommy Emerson

Chair

Board of County Commissioners of

Chatham County

186 Emerson Road

Siler City, NC 27344

Kelvin Henderson

Plant General Manager

Carolina Power and Light Company

Shearon Harris Nuclear Power Plant

Electronic Mail Distribution

cc w/encl. (continued page 4)

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

05000400/2009006

Distribution w/encl:

C. Evans, RII EICS

L. Slack, RII EICS

OE Mail

RIDSNRRDIRS

PUBLIC

RidsNrrPMShearonHarris Resource

Enclosure

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

3

Enclosure

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

4

Enclosure

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

annunciators.

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

Attachment

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

3

Attachment

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

4

Attachment

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

5

Attachment

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

6

Attachment

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