ML993540229

IRC Harris 1999002 Integrated

ML993540229
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Issue date:	05/21/1999
From:	Division Reactor Projects II
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May 21, 1999 Carolina Power & Light Company ATTN: Mr. James Scarola Vice President - Harris Plant Shearon Harris Nuclear Power Plant P. O. Box 165, Mail Code: Zone 1 New Hill, NC 27562-0165

SUBJECT:

NRC INTEGRATED INSPECTION REPORT 50-400/99-02

Dear Mr. Scarola:

On April 24, 1999, the NRC completed an inspection at your Shearon Harris facility. The enclosed report presents the results of that inspection.

During the six-weeks covered by this inspection period, our inspectors found that your staff generally took a safety-conscious approach to the operation of the facility.

Based on the results of this inspection, the NRC has determined that three violations of NRC requirements occurred. These violations are being treated as Non-Cited Violations (NCVs),

consistent with Appendix C of the Enforcement Policy. These NCVs are described in the subject inspection report. If you contest the violation or severity level of these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001, with copies to the Regional Administrator, Region II; the NRC Resident Inspector at Shearon Harris Nuclear Power Plant; and the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001.

During the inspection period, the NRC received a response, dated March 31, 1999, to a Notice of Violation 50-400/98-11-01, 50-400/98-11-02, and 50-400/98-11-03. We have evaluated your response, and found that it meets the requirements of 10 CFR 2.201.

CP&L 2

In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and its enclosures will be placed in the NRC Public Document Room (PDR).

Sincerely, (Original signed by G. T. MacDonald)

Brian R. Bonser, Chief Reactor Projects Branch 4 Division of Reactor Projects Docket No.:

50-400 License No.: NPF-63

Enclosure:

Inspection Report cc w/encl: (See page 3)

CP&L 3

cc w\\encl:

Terry C. Morton, Manager Performance Evaluation and Regulatory Affairs CPB 9 Carolina Power & Light Company P. O. Box 1551 Raleigh, NC 27602-1551 Chris L. Burton Director of Site Operations MC: Zone 1 Carolina Power & Light Company Shearon Harris Nuclear Power Plant P. O. Box 165 New Hill, NC 27562-0165 Bo Clark Plant General Manager--Harris Plant Carolina Power & Light Company Shearon Harris Nuclear Power Plant P. O. Box 165 New Hill, NC 27562-0165 Donna B. Alexander, Manager Regulatory Affairs Carolina Power & Light Company Shearon Harris Nuclear Power Plant P. O. Box 165, Mail Zone 1 New Hill, NC 27562-0165 Johnny H. Eads, Supervisor Licensing/Regulatory Programs Carolina Power & Light Company Shearon Harris Nuclear Power Plant P. O. Box 165, Mail Zone 1 New Hill, NC 27562-0165 William D. Johnson Vice President & Corporate Secretary Carolina Power & Light Company P. O. Box 1551 Raleigh, NC 27602 John H. O'Neill, Jr.

Shaw, Pittman, Potts & Trowbridge 2300 N. Street, NW Washington, DC 20037-1128 Mel Fry, Director Division of Radiation Protection N. C. Department of Environmental Commerce & Natural Resources 3825 Barrett Drive Raleigh, NC 27609-7721 Karen E. Long Assistant Attorney General State of North Carolina P. O. Box 629 Raleigh, NC 27602 Public Service Commission State of South Carolina P. O. Box 11649 Columbia, SC 29211 Chairman of the North Carolina Utilities Commission P. O. Box 29510 Raleigh, NC 27626-0510 Robert P. Gruber Executive Director Public Staff NCUC P. O. Box 29520 Raleigh, NC 27626 Vernon Malone, Chairman Board of County Commissioners of Wake County P. O. Box 550 Raleigh, NC 27602 Richard H. Givens, Chairman Board of County Commissioners of Chatham County P. O. Box 87 Pittsboro, NC 27312 Distribution w/encl: (See page 4)

CP&L 4

Distribution w/encl:

L. Plisco, RII B. Bonser, RII R. Laufer, NRR G. MacDonald, RII PUBLIC NRC Resident Inspector U. S. Nuclear Regulatory Commission 5421 Shearon Harris Road New Hill, NC 27562-9998 OFFICE DRP/RII DRP/RII DRP/RII DRP/RII DRP/RII DRS/RII EICS/RII SIGNATURE NAME FJape JBrady BHagar MShannon CCasto FWright ABoland DATE 6/ /25 6/ /25 6/ /25 6/ /25 6/ /25 6/ /25 6/ /25 COPY?

YES NO YES NO YES NO YES NO YES NO YES NO YES NO OFFICIAL RECORD COPY DOCUMENT NAME: G:\\HARRIS\\REPORT\\9902.drp

Enclosure U. S. NUCLEAR REGULATORY COMMISSION REGION II Docket No:

50-400 License No:

NPF-63 Report No:

50-400/99-02 Licensee:

Carolina Power & Light (CP&L)

Facility:

Shearon Harris Nuclear Power Plant, Unit 1 Location:

5413 Shearon Harris Road New Hill, NC 27562 Dates:

March 14 - April 24, 1998 Inspectors:

J. Brady, Senior Resident Inspector R. Hagar, Resident Inspector M. Shannon, Senior Resident Inspector, Sequoyah (Sections O2.1 and M7.1)

F. Wright, Senior Radiation Specialist (Sections R1.1, R1.2, R2.1 and R7.1)

F. Jape, Senior Project Manager (Section E8.3)

Approved by:

B. Bonser, Chief, Projects Branch 4 Division of Reactor Projects

EXECUTIVE

SUMMARY

Shearon Harris Nuclear Power Plant, Unit 1 NRC Inspection Report 50-400/99-02 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a six-week period of resident inspection.

Operations During the period, the conduct of operations was in accordance with applicable procedures (Section O1.1).

A violation was identified for failure to comply with a Technical Specification (TS) action statement while testing reactor trip breakers in Mode 3. The new procedure prepared to test the breaker in mode 3 did not include guidance addressing applicable TS action statements. Additionally, operators did not independently identify the TS action statements in a timely manner (Section O1.2).

Operators successfully completed a reactor startup and a subsequent synchronization of the main generator to the grid. Procedures were correctly used, communications were effective, and operator actions were safety conscious. (Section O1.3).

The Component Cooling Water system was operable and being maintained in an acceptable manner (Section O2.1).

Plant Nuclear Safety Committee activities related to a reactor trip were effective in ensuring that the root cause was identified and the necessary corrective actions were taken before the unit was restarted (Section O7.1).

Maintenance An unresolved item was opened for issues related to a containment isolation valve which failed to close in response to a signal from a slave relay (Section M1.1).

The surveillance performances were conducted in accordance with applicable procedures (Section M2.1).

A violation was identified for failure to calibrate component cooling water (CCW) instrumentation required for inservice testing. In addition, a negative finding was identified for not including CCW control room alarm response instrumentation in a routine calibration program (Section M7.1).

2 Engineering The investigation and reportability evaluation, related to a containment isolation valve which failed to close in response to an actuation signal from a slave relay, was not thorough (Section E7.1).

The Y2K checklist, per TI 2515/141, was completed. Overall, the Y2K project is about 90 percent complete and the contingency plan is about 80 percent complete (Section E8.3).

Plant Support Radiation Protection, Plant Security, and Fire Protection activities were accomplished in accordance with applicable site procedures. (Sections R1, S1, and F1)

Licensee radiation surveys, postings, access controls, and radiological work controls were effective and performed in accordance with regulatory requirements. (Section R1.1)

The Harris ALARA program was effective in continued reduction of site collective personnel radiation doses.

(Section R1.2)

Reviewed radiation monitoring instrumentation calibration documentation was in order and completed in accordance with licensee procedures. Radiation monitoring instruments were operable and had valid calibration certifications. (Section R2.1)

The E&RC staff personnel interviewed understood the value of effective corrective action processes and good utilization of those processes were observed. The documentation and corrective actions for E&RC program condition reports were generally good. (Section R7.1)

Report Details Summary of Plant Status Unit 1 began this inspection period with the unit shutdown, following a March 12 reactor trip.

The unit was restarted on March 19, and remained at 100% power for the remainder of the inspection period.

I. Operations O1 Conduct of Operations O1.1 General Comments

a.

Inspection Scope (71707)

The inspectors conducted frequent reviews of ongoing plant operations including control room tours, shift turnovers, and observation of operations surveillance activities. The inspectors also conducted frequent tours of the facility to verify equipment condition, housekeeping, and proper use of clearances.

b.

Observations and Findings In general, the conduct of operations was professional and safety-conscious. Routine activities were adequately performed. Operations shift crews were appropriately sensitive to plant equipment conditions and maintained a questioning attitude in relation to unexpected equipment responses. Facilities and equipment were maintained, and clearances were installed and removed in accordance with applicable procedures.

c.

Conclusions During the period, the conduct of operations was in accordance with applicable procedures.

O1.2 Non-compliance with Technical Specification 3.3.2

a.

Inspection Scope (71707)

The inspectors reviewed the circumstances associated with testing a reactor trip breaker on March 14.

2

b.

Observations and Findings During the shutdown following the March 12 reactor trip, while the unit was in Technical Specification (TS) Mode 3, the licensee decided to replace the reactor trip breaker that had failed its surveillance test on February 17. (That failure was described in Section O1.2 of NRC Inspection Report (IR) 50-400/99-01).

On March 13, in order to replace and perform post-maintenance testing on the trip breaker, the licensee placed one train of the Solid State Protection System (SSPS) in the test mode. That action rendered inoperable one channel of several SSPS functional units. Consequently, several action statements in TS 3.3.2 became applicable, including action statement 21, which was associated with the automatic actuation logic and actuation relays for main steam line actuation and auxiliary feedwater. That action statement required either restoration of operability or entry into TS Mode 4 within six hours. However, as reported in LER 50-400/1999-005-00, the licensee did not comply with that action statement. Instead, the licensee left one train of SSPS inoperable for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and 28 minutes while the unit remained in Mode 3. The licensee was thus out of compliance with TS 3.3.2 for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 28 minutes. The inspectors determined that this noncompliance occurred for the following reasons:

Because the licensee did not have a procedure for performing this evolution in Mode 3, they developed one. The new procedure, based on the procedure for performing this evolution in Mode 5, did not identify all of the TS action statements that would be applicable in Mode 3. Specifically, the procedure did not identify that TS 3.3.2 action statement 21 would be applicable.

The operators did not independently and in a timely manner identify the TS action statements that would become applicable by placing an SSPS train into the test mode. The operators eventually did identify the applicable action statements, but by the time they did so and restored operability, the affected SSPS train had been in the test mode for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and 28 minutes.

The inspectors noted that the licensee initiated condition report (CR) 99-00735 to document this event.

TS 3.3.2 requires for automatic actuation logic and actuation relays for main steam line actuation and auxiliary feedwater that two channels be operable in Modes 1,2, and 3; if only one channel is operable, action statement 21 requires, in part, that the unit be placed in hot standby within six hours. The inspectors thus considered the licensees failure to comply with action statement 21 on March 13 to be a violation of TS 3.3.2.

The inspectors noted that the licensee restored compliance within a reasonable time after the violation was identified, and placed the violation into the corrective action program to prevent recurrence. Furthermore, the inspectors considered that the violation was not repetitive as a result of inadequate corrective action, and was not willful. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. This violation is in the

3 licensees corrective action program as CR 99-00735. The inspectors have designated this violation as NCV 50-400/99-02-01, failure to comply with TS 3.3.2.

c.

Conclusions A violation was identified for failure to comply with a TS action statement while testing reactor trip breakers in Mode 3. The new procedure prepared to test the breaker in mode 3 did not include guidance addressing applicable TS action statements.

Additionally, operators did not independently identify the TS action statements in a timely manner.

O1.3 Plant Startup on March 19 (71707)

The inspectors observed operator performance during the reactor startup, the increase in reactor power to approximately 15% of rated thermal power, and the subsequent synchronization of the main generator to the grid on March 19. The inspectors observed that the operators carefully followed procedures, and employed effective communications techniques throughout the evolution. The operators responded conservatively to an unexpected indication on the main control board, by stopping the evolution until Engineering verified that no problem existed. The operators increased power cautiously, and after verifying that plant conditions satisfied procedural requirements, the operators successfully completed synchronizing the main generator to the grid.

O2 Operational Status of Facilities and Equipment O2.1 Component Cooling Water (CCW) System Walkdown

a.

Inspection Scope (71707)

The inspectors performed a detailed walkdown of the accessible mechanical and electrical portions of the CCW system; reviewed condition reports, corrective work history, related TS and Final Safety Analysis Report (FSAR) sections, the system description, surveillance records, calibration records, and trends in pump and valve inservice testing in accordance with Section XI of the ASME Boiler & Pressure Vessel Code. The inspectors also discussed system operation and testing with the system engineer and the Section XI pump and valve test engineer.

b.

Observations and Findings During the walkdown, the inspectors verified that the accessible valves in the main flow paths with visually verifiable positions were correctly positioned, locked as specified, and throttled to the required positions consistent with the applicable system drawings and lineup procedures. The inspectors also verified correct electrical breaker positions for selected equipment. The inspectors verified valve handwheels in place and observed no instances of bent valve stems, significant system or component leaks, corroded components, or adverse environmental conditions. The inspectors examined selected

4 local and remote indicators, motor operators, piping supports, and system instrumentation and found no undocumented material issues or deficiencies. The inspectors observed that the CCW system components and piping were clearly labeled with easily readable component identification tags and system flow paths. Spaces were found to be very clean and free of debris, loose materials, ignition sources, flammable materials, and ancillary equipment interference. The inspectors also verified that the various automatic system actuations were routinely tested, and that the testing verified that the system would perform its automatic function. In addition, the inspectors reviewed the licensees maintenance rule event log report for the CCW system, to evaluate the disposition of functional failures and determined that the functional failures had been appropriately considered under the maintenance rule.

The inspectors identified various deficient conditions during the inspection. These deficient conditions included failures to calibrate instrumentation used to perform surveillance testing required by ASME Section XI and not including instrumentation used to provide control room alarms in a calibration program (Section M7.1) and an inadequate post-maintenance test revealed by an inoperable containment isolation valve (Section M1.1).

c.

Conclusions Based upon a detailed walkdown of accessible portions of the CCW system, the inspectors concluded that the system was operable and being maintained in an acceptable manner.

O4 Operator Knowledge and Performance O4.1 General Comments (71707)

Throughout the inspection period, the inspectors observed operator performance in a variety of circumstances. In those circumstances, the inspectors found that both licensed and non-licensed operators participated attentively in pre-job briefs, followed procedural instructions, and used effective place-keeping techniques. The inspectors also observed that the operators effectively used three-way communication techniques.

The inspectors concluded that operator knowledge and performance was in accordance with applicable procedural requirements, except as noted above in Section O1.2.

O7 Quality Assurance in Operations O7.1 General Comments

a.

Inspection Scope (71707, 93702)

During the inspection period, the inspectors reviewed multiple licensee quality assurance activities, including:

Condition Reports; and

5 Plant Nuclear Safety Committee (PNSC) meetings on March 14 and March 15.

b.

Observations and Findings The inspectors found that the licensee promptly and consistently initiated condition reports for identified adverse conditions. The PNSC discussions on March 14 and March 15 focused on issues associated with the March 12 reactor trip (discussed in Section O1.4 of NRC Inspection Report 50-400/99-01). The inspectors observed that the scope of those discussions included the major issues that had been associated with the trip and that the PNSC actively reviewed and considered those issues. The inspectors observed that as various staff members presented reports to the PNSC, PNSC members questioned and challenged those presenters. The inspectors concluded that the PNSC critically reviewed the presented reports.

c.

Conclusions PNSC activities related to a reactor trip were effective in ensuring that the root cause was identified and the necessary corrective actions were taken before the unit was restarted.

O8 Miscellaneous Operations Issues (92700, 92901)

O8.1 Closure of Open Severity Level IV Violations The NRC recently revised NUREG-1600, Rev. 1, General Statement of Policy and Procedures for NRC Enforcement Actions, (Enforcement Policy) by the addition of Appendix C. Appendix C, Interim Enforcement Policy for Power Reactor Severity Level IV Violations, effective March 11, 1999, revises the NRCs enforcement approach for Severity Level IV violations. Appendix C permits closure of most Severity Level IV violations, based on the violation being entered into the licensees corrective action program, as well as other considerations as described in the Appendix. The NRC has conducted a review of the following Severity Level IV violations, and considers it appropriate to close these violations consistent with Appendix C of the Enforcement Policy:

Violation Number Corresponding Condition Report Number(s) 50-400/98-01-01 98-01451 98-00592 98-02318 98-03267 50-400/98-01-04 98-00428 50-400/98-06-02 98-01014 50-400/98-08-03 99-08037

6 50-400/98-09-02 98-02832 50-400/98-11-01 99-00576 50-400/98-11-02 98-03264 99-00574 O8.2 (OPEN) LER 50-400/1999-005-01: Engineered Safety Features Actuation Systems Technical Specifications exceeded. This event was described above in Section O1.2.

This LER remains open pending inspector verification of licensee corrective actions.

II. Maintenance M1 Conduct of Maintenance M1.1 Inoperable Containment Isolation Valve

a.

Inspection Scope (62707, 61726)

The inspectors reviewed the circumstances and corrective actions related to a containment-isolation valve which failed to stroke closed during a TS required surveillance test.

b.

Observations and Findings The inspectors reviewed the circumstances associated with CR 98-03211, which documented that on December 6, 1998, containment isolation valve 1CC-176, component cooling water to the reactor coolant drain tank and excess letdown heat exchanger, failed to stroke closed during the performance of OST-1045, ESFAS Train B Slave Relay Test Quarterly Interval Modes 1-4," Revision 14. The inspectors review of CR 98-03211 is discussed in Section E7.1 of this report. Subsequent troubleshooting found that within the Limitorque operator on the valve, a finger base contact was misaligned and a contact lead was loose. After those problems were corrected, the valve operated properly. The inspectors found later that the misaligned finger base contact and loose contact lead were both in the circuit that actuated the valve in response to a signal from the Engineered Safeguards Features Actuation System (ESFAS) slave relay.

Discussions with the system engineer indicated that preventive maintenance had been performed on the valve operator on November 11, 1998, in accordance with procedure PM-M-0014, Limitorque Inspection and Lubrication, Revision 15. The inspectors found that following the November 11 maintenance, post-maintenance testing had been performed in accordance with Equipment Inoperable Record (EIR) 98-752T, and that the only testing specified on that EIR was to cycle valve twice (from the handswitch in the

7 main control room). The inspectors noted that the circuit used to actuate the valve from the handswitch was different from the circuit used to actuate the valve from the ESFAS slave relay. The inspectors therefore concluded that cycling 1CC-176 from the control room handswitch did not demonstrate that 1CC-176 would actuate in response to a signal from the ESFAS slave relay, and that its operability was therefore indeterminate.

The inspectors concluded that if any valve was subjected to PM-M-0014 maintenance and post-maintenance testing that included only cycling the valve from the control room handswitch, then the operability of that valve with respect to its response to a signal from an ESFAS slave relay would be indeterminate.

The inspectors noted that preventive maintenance in accordance with PM-M-0014 had been and was being regularly performed on all safety-related valves with Limitorque operators. Discussions with Work Control Center (WCC) personnel revealed that WCC staff routinely require only cycling the valve twice (from the control room handswitch) after PM-M-0014 work is completed. The inspectors questioned whether the licensee had demonstrated that other valves subject to PM-M-0014 maintenance, were in fact operable with respect to responding to a signal from an ESFAS slave relay. In response to the inspectors questions, the licensee identified the safety-related valves in that category, reviewed the maintenance and testing history of those valves, and determined that since the last performance of PM-M-0014 on those valves, each valve had been successfully actuated in response to a slave-relay signal during surveillance testing. In addition, the licensee placed PM-M-0014 on administrative hold. The inspectors considered the licensees response to this concern to be timely and thorough.

From a review of procedure OMM-014, Operation of the Work Control Center, Revision 18, the inspectors determined that EIR 98-752T had been prepared in accordance with procedural guidance. The inspectors observed that although Section 5.1.1 (Standard Practices) of that procedure provided guidance that manually operated Limitorque motor operated valves are required to be stroked electrically from the control room switch to be declared operable, that section does not provide similar guidance for automatically-actuated valves.

As discussed above, because the November 11 post-maintenance test did not demonstrate that 1CC-176 would perform satisfactorily in service with respect to slave-relay actuation, the operability of 1CC-176 following that test was indeterminate. The inspectors therefore questioned whether 1CC-176 had been operable when the unit entered Mode 4, as required by TS 3.6.3. In response to the inspectors questions, the licensee initiated an investigation. Pending completion of that investigation, the inspectors review of the investigation results, and the subsequent completion of the inspectors assessment of the safety and risk significance of this event, this issue has been designated as Unresolved Item (URI) 50-400/99-02-02, failure of a containment isolation valve to close in response to a slave-relay signal.

c.

Conclusions An unresolved item was opened for issues related to a containment isolation valve which failed to close in response to a signal from a slave relay.

8 M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Surveillance Observation (61726)

The inspectors observed all or portions of the following surveillance tests:

Test Title Revision MST-I0128 Main Steamline Pressure, Loop 2 (P-0484) Operational Test 6

OST-1005 Control Rod and Rod Position Indicator Exercise Monthly Interval Modes 1-3 8

OST-1094 Sequencer Block Circuit and Containment Fan Cooler Testing Train A Quarterly Interval All Modes 5

The inspectors found that the testing was adequately performed and conducted in accordance with applicable procedures.

M7 Quality Assurance in Maintenance M7.1 Failure to Calibrate Instrumentation Used for Inservice Testing (IST)

a.

Inspection Scope (61726)

The inspectors reviewed the implementation of the instrument calibration program for instruments used on the CCW system.

b.

Observations and Findings During a detailed CCW system walkdown, the inspectors reviewed the instrument calibration program to ensure that the appropriate CCW instrumentation was incorporated into the program and had been calibrated as specified. The inspectors noted that the IST program for the CCW pumps used the installed CCW pump suction and discharge pressure instruments to measure pump performance. However, the inspectors noted that these instruments were not included in the licensees calibration program and had not been calibrated since 1995. 10 CFR 50, Appendix B, Criterion XII, Control of Measuring and Test Equipment, requires, in part, that measures shall be established to assure that... instruments... used in activities affecting quality are properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within necessary limits. ASME/ANSI Oma-1998, Part 6, Inservice Testing of Pumps in Light Water Reactor Power Plants, requires that, Instruments...shall be calibrated...

The inspectors considered the failure to include the CCW pump suction and discharge pressure instruments in a periodic calibration program to be a violation of 10 CFR 50, Appendix B, Criterion XII. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. This

9 violation is in the licensees corrective action program as CR 99-00716. The inspectors have identified this violation as NCV 50-400/99-02-03, failure to calibrate instrumentation required for inservice testing.

In addition, the inspectors identified other instruments that were not in a routine calibration program. On March 10, during a review of CCW system related control room alarm response procedures, the inspectors noted that two CCW system instruments that provide an input to specific control room alarms were not included in the licensees routine calibration program. The first instrument was for the CCW flow to the B spent fuel pool cooling heat exchanger (tag number FIS-04CC-0640CS) used in alarm response procedure ALB-005-3-5A and 5B, Spent Fuel HX B CCW Low Flow, and Spent Fuel HX B CCW High Flow. The second instrument was for the CCW return header temperature from the combined reactor coolant pump thermal barriers (tag number TIS-01CC-0684W) used in alarm response procedure ALB-005-2-2B, RCP Thermal Barrier Outlet High Temperature. This issue was discussed with the system engineer and condition report CR 99-00707 was initiated on March 11. Work orders (99-ACSK1 and 99-ACSI1) were initiated to calibrate the equipment immediately.

c.

Conclusions A violation was identified for failure to calibrate CCW instrumentation required for inservice testing. In addition, a negative finding was identified for not including CCW control room alarm response instrumentation in a routine calibration program.

M8 Miscellaneous Maintenance Issues (92700)

M8.1 (Closed) LER 50-400/97-021-00, -01, -02, -03: Technical Specification Surveillance Procedure Review Project Identified Deficiencies. The programmatic nature of the surveillance procedure program deficiencies was addressed from an enforcement perspective in NRC letter to CP&L dated August 29, 1997. That letter described the bases for applying enforcement discretion to the numerous surveillance procedure deficiencies that had been identified up to that time. The bases included comprehensive corrective actions have been implemented for identified deficiencies and further review of TS surveillance procedures is planned. The surveillance procedure deficiencies described in these LERs are part of the further reviews discussed in the August 29, 1997 letter.

The reported violations of TS 4.9.11 related to spent fuel pool level, TS 4.9.12.c related to Fuel Handling Building Emergency Exhaust System charcoal testing, TS 4.7.6.c related to Reactor Auxiliary Building Emergency Exhaust System (RABEES) charcoal testing, and TS 4.4.4.1.b related to testing of valve 1RC-116 were caused by inadequate surveillance procedures. TS 6.8.1.a and Regulatory Guide 1.33, Appendix A, Section 8.b requires that written procedures shall be established implemented and maintained for surveillance tests listed in the Technical Specifications. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. These violation examples are in the licensees corrective action program as condition reports 97-03863, 97-4227, 97-4228, 98-00463. The

10 inspectors have designated these violations as examples 1 through 4 of NCV 50-400/99-02-04, inadequate surveillance procedures.

The NRC review of the follow-on surveillance procedure review project was documented in IR 50-400/98-06, Section M8.6. The inspectors reviewed the specific corrective actions for the items reported in the LERs and found them to have been adequately implemented. This included changing operator logs and submitting a Technical Specification change (September 1, 1998) for spent fuel pool water level; implementing temporary actions for Fuel Handling Building Emergency Exhaust System bleed flow and installing a permanent modification (ESR 97-00737) to remove the bleed flow (amendment 82 to the Technical Specifications removed the bleed flow measurement requirement from Section 4.9.12); performing an engineering evaluation (ESR 97-00700) for the RABEES that determined bleed flow had no effect due to the bleed flow location; and revising procedure OST 1117, Pressurizer Safety Grade PORV Operability -

Quarterly Test to include testing of valve 1RC-116, Revision 6.

M8.2 (Closed) LER 50-400/98-006-00: Failure to perform inspections and preventive maintenance on molded case circuit breakers as required by Technical Specifications.

This item involved inadequate surveillance procedures for testing of 480 volt molded case circuit breakers in accordance with TS 4.8.4.1.b. and had not yet been reviewed by the Technical Specification Surveillance Procedure Review Project when it was found by the licensees line organization. The inspectors considered this violation an additional example (5) of NCV 50-400/99-02-04, inadequate surveillance procedures. This violation example is in the licensees corrective action program as CR 98-00931.

Corrective action included testing of the subject breakers ( 9 pressurizer heater supply breakers) on May 8, 1998 with no deficiencies identified. The surveillance procedures were revised to include the manufacturer requirement to cycle the breakers once every 60 months.

III. Engineering E7 Quality Assurance in Engineering E7.1 Condition Report Review

a.

Inspection Scope (37551)

The inspectors reviewed the Adverse Condition Evaluation report prepared by Engineering for CR 98-03211. This CR described the failure of a containment isolation valve to stroke closed during a TS required surveillance test. That failure is discussed in Section M1.1 of this report.

b.

Observations and Findings The inspectors observed that CR 98-03211 had been classified as an Adverse Condition, and that the evaluating supervisor had determined that a low-level

11 investigation was appropriate. The inspectors noted that for a CR classified this way, procedure AP-605, Condition Report Evaluations, Revision 17, requires only that the investigator gather as much information as feasible to determine the most likely or apparent cause of the event/issue and document the findings; it does not require a detailed root-cause investigation. The CR described the event as a failure of the subject valve to shut from its slave relay on December 6, 1998, and the report stated that the cause of the event was insufficient detail or guidance in the procedure that was followed to perform preventive maintenance on the valve on November 11, 1998. The adverse condition evaluation report identified that the problem with the valve actuator most likely occurred during this preventive maintenance activity.

During the review of the CR, the inspectors found:

Although the problem most likely occurred during the preventive maintenance activity, the licensee did not address this in the corrective action program as a rework issue in accordance with plant procedures.

Tthe post-maintenance test performed on the valve actuator was not adequate to reveal improperly-performed maintenance (as discussed above in Section M1.1),

and The valve may have been inoperable during the entire period from the completion of the November 11 preventive-maintenance activity to the completion of corrective maintenance following the December 6 failure.

However, the inspectors noted that the report did not adequately address any of the above findings. The inspectors therefore concluded that the investigation described in the report was not thorough.

The inspectors consideration of the initiation, characterization, and disposition of this CR will be included in URI 50-400/99-02-02. That URI is discussed in Section M1.1.

c.

Conclusions The investigation and reportability evaluation, related to a containment isolation valve which failed to close in response to an actuation signal from a slave relay, was not thorough.

E8 Miscellaneous Engineering Issues (92700, TI 2515/141)

E8.1 (Closed) LER 50-400/98-001-00: Potential Condition Outside Design Basis related to Instrument Air System Leak causing the S/G Pre-Heater Bypass Isolation Valves to be Inoperable. This LER reported Violation 50-400/97-13-02 which was closed in IR 50-400/98-10.

12 E8.2 (Closed) VIO 50-400/98-11-03: inadequate design of the reactor coolant system with the vacuum skid attached. The inspectors reviewed the licensees response dated March 31, 1999, and noted the corrective actions described therein. The inspectors verified that the actions described in the licensees response have been completed, and consider the problem described in the violation to be corrected.

E8.3 Year 2000 (Y2K) Readiness Program Review The staff conducted an abbreviated review of Y2K activities and documentation using Temporary Instruction (TI) 2515/141, Review of Year 2000 (Y2K) Readiness of Computer Systems at Nuclear Power Plants.

The review addressed aspects of Y2K management planning, documentation, implementation planning, initial assessment, detailed assessment, remediation activities, Y2K testing and validation, notification activities, and contingency planning. The reviewers used NEI/NUSMG 97-07, Nuclear Utility Year 2000 Readiness, and NEI/NUSMG 98-07, Nuclear Utility Year 2000 Readiness Contingency Planning, as the primary references for this review.

During the review, the licensee stated that the Y2K Readiness Project activities were 90% completed with contingency planning being approximately 80% complete, and that both programs were on target to be completed by their scheduled due dates.

Conclusions regarding the Y2K readiness of the facility are not included in this report.

The results of this review will be combined with the results of reviews of other licensees in a summary report to be issued by July 31, 1999.

IV. Plant Support R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Radiological Work Controls

a.

Inspection Scope (83750)

The inspectors reviewed radiation protection (RP) activities against applicable RP program requirements and 10 CFR Part 20. The inspection included reviews of records and procedures, interviews with licensee personnel, and observations of work activities in progress. The inspectors made observations in the fuel handling, radioactive waste processing, and reactor auxiliary buildings.

b.

Observations and Findings Independent radiation surveys made by the inspector were in agreement with licensees radiation survey results. The radiological postings were adequate for areas surveyed.

Locked high radiation doors checked by the inspectors were secured properly.

13 There was good RP coverage at the main radiological control area entrance and exit portals, and the inspectors observed good interactions and communications between radiation workers and RP personnel.

The inspectors attended a pre-job briefing during the inspection for work planned in a high radiation area. The inspectors noted excellent participation by the radiation workers assigned to the task and good exchange of information between the radiation workers and the RP personnel.

Individual occupational radiation worker doses remained low, with the highest total effective dose equivalent well below the licensees 2,000 millirem per year administrative limit.

c.

Conclusions Licensee radiation surveys, postings, access controls, and radiological work controls were effective and performed in accordance with regulatory requirements.

R1.2 As Low As Reasonably Achievable (ALARA)

a.

Inspection Scope (83750)

Implementation of the ALARA program for Re-Fueling Outage cycle 8 (RFO-8) activities was evaluated.

b. Observations and Findings Licensee goals for the sites RFO-8 included, in part, personnel safety goals, a duration goal of less than 38 days and, a collective radiation dose goal less than 101.35 person-rem. The collective dose goals and actual dose received are shown in the table below.

In 1998 the licensee met the non-outage dose goal, but failed to meet the annual and outage dose goals.

1998 Harris Collective Personnel Exposure Goals Site Annual Person-Rem Non-Outage Person-Rem RFO-8 Person-Rem Length (days)

Goals 128.8 10.4 101.5 38 Actual 133.4 8.5 117.6 35 The 1998 annual dose goal would have been achieved if the refueling dose goal had been met. Activities contributing to additional outage dose included expanding valve work, and problems with a reactor head seal weld and scaffolding. The RFO-8 dose like the RFO-7 dose in 1997 was still lower than previous RFOs. A level 1 condition

14 report was initiated by the licensee to identify and correct scaffolding problems and the ALARA personnel planned to be more involved in the planning process to improve the accuracy of dose projection. The licensee continued to reduce collective dose as shown below.

Annual Collective Doses (Person Rem)

Year 1995 1996*

1997 1998 1999*

Dose 174 17 149 134

<18.5 GOAL

• Years without RFO Chemistry control procedures were used to reduce primary system contamination levels and dose rates during planned reactor shutdowns. Licensee documentation of reactor coolant system radioactivity for RFO-8 indicated the shutdown procedures effectively removed contamination from the primary system and compared well with previous system cleanups.

c.

Conclusions The Harris ALARA program was effective in continued reduction of site collective personnel radiation doses.

R2.1 Radiological Survey and Monitoring Instrumentation a

Inspection Scope (83750)

The maintenance and calibration of radiation survey and monitoring instruments were evaluated. Calibration procedures, calibration records, and traceability of calibration standards were reviewed for selective instruments.

b.

Observations and Findings Specific radiation detection instruments including portable low and high range ion-chambers, Geiger Muller radiation survey instruments, low level gamma scintillation detectors, personnel contamination friskers, and continuous air monitors were selected for the review. All instruments were calibrated in accordance with applicable procedures and all calibration documentation was in order. During plant tours the inspectors verified that radiation monitoring instrumentation was operational and had valid calibration certifications.

c.

Conclusions Radiation monitoring instrumentation calibration documentation reviewed was in order and completed in accordance with licensee procedures. Radiation monitoring

15 instruments were operable and had valid calibration certifications.

R7 Quality Assurance In RP&C Activities R7.1 Documentation and Corrective Actions

a.

Inspection Scope (83750)

Adverse condition reports and feed back reports for the Environmental & Radiation Control (E&RC) department since the previous inspection (November of 1998) were reviewed.

b.

Observations and Findings The inspectors reviewed recent RP issues identified in the licensees corrective action system. The inspectors determined the E&RCs threshold for placing issues into the licensees corrective action program appeared to be low for regulatory compliance issues and appropriate to make program improvements and meet E&RC goals. The reviewed CRs included good analysis of problems with appropriate corrective actions to prevent recurrence. For the period reviewed, approximately 160 items were documented in the system. The E&RC group was self-identifying approximately 85 percent of the CRs in the system. The inspectors also attended an E&RC CR meeting held to discuss progress of open condition reports and to make assignments for new CRs. The inspectors verified that the licensee was trending problems identified by the staff and discussed those trends with the E&RC department manager. No significant adverse trends were identified. In interviews with various E&RC staff the inspectors found they understood the value of an effective corrective action process.

c.

Conclusions The E&RC personnel interviewed understood the value of effective corrective action processes and good utilization of those processes were observed. The documentation and corrective actions for E&RC program condition reports were generally good.

S1 Conduct of Security and Safeguards Activities S1.1 General Comments (71750)

The inspectors observed security and safeguards features and activities during the conduct of plant tours, including:

general integrity of the protected area barrier maintenance of the isolation zones illumination levels access control vital area controls The inspectors concluded that security and safeguards activities were conducted in

16 accordance with applicable procedures and the Security Plan.

F1 Control of Fire Protection Activities F1.1 General Comments (71750)

During the conduct of tours and observation of maintenance activities, the inspectors observed fire protection equipment and activities, and found that equipment and those activities to be in accordance with applicable plant procedures.

V. Management Meetings X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on April 23, 1999. The licensee acknowledged the findings presented.

The inspectors asked the licensee whether any of the material examined during the inspection should be considered proprietary. No proprietary information was identified.

17 PARTIAL LIST OF PERSONS CONTACTED Licensee D. Alexander, Manager, Regulatory Affairs J. Bates, Superintendent, Environmental and Chemistry D. Batton, Superintendent, On-Line Scheduling D. Braund, Superintendent, Security B. Clark, General Manager, Harris Plant A. Cockerill, Superintendent, I&C Electrical Systems J. Cook, Manager, Outage and Scheduling J. Eads, Supervisor, Licensing and Regulatory Programs R. Field, Manager, Nuclear Assessment T. Hobbs, Acting Manager, Operations M. Keef, Manager, Training G. Kline, Manager, Harris Engineering Support Services K. Neuschaefer, Manager, Environmental & Radiation Control T. Pilo, Superintendent, Radiation Protection J. Scarola, Vice President, Harris Plant B. Waldrep, Manager, Maintenance NRC B. Bonser, Chief, Reactor Projects Branch 4 R. Laufer, Harris Project Manager, NRR INSPECTION PROCEDURES USED IP 37551:

Onsite Engineering IP 61726:

Surveillance Observations IP 62707:

Maintenance Observation IP 71707:

Plant Operations IP 71750:

Plant Support Activities IP 83750:

Occupational Radiation Exposure IP 92700:

Onsite Followup of Events IP 92901:

Followup - Plant Operations IP 93702:

Onsite Response to Events TI 2515/141: Review of Year 2000 (Y2K) Readiness of Computer Systems at Nuclear Power Plants

18 ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-400/99-02-01 NCV Failure to Comply With TS 3.3.2 (Section O1.2).

50-400/99-02-03 NCV Failure to Calibrate Instrumentation Required for Inservice Testing (Section M7.1).

50-400/99-02-04 NCV Inadequate Surveillance Procedures (5 examples)

(Sections M8.1 and M8.2).

50-400/99-02-02 URI Failure of a Containment Isolation Valve to Close in Response to a Slave-Relay Signal (Section M1.1).

Closed 50-400/98-01-01 VIO Failure to Follow Procedures (Section O8.1).

50-400/98-01-04 VIO Failure to Properly Implement and Maintain the Applicable Fire Protection Program Design Control Documentation Requirements for Fire Barrier Penetration Seals (Section O8.1).

50-400/98-06-02 VIO Performance of During Shutdown Surveillance at Power (Section O8.1).

50-400/98-08-03 VIO Failure to Document Incorrect Safety Evaluation as an Adverse Condition (Section O8.1).

50-400/98-09-02 VIO Failure to Translate Design Requirements Into Maintenance Procedures for Agastat E7000 Series Relays (Section O8.1).

50-400/98-11-01 VIO Failure to Promptly Identify and Correct a Condition Adverse to Quality (Section O8.1).

50-400/98-11-02 VIO Failure to Effectively Implement the Post-Trip Review Procedure (Section O8.1).

50-400/98-11-03 VIO Inadequate Design of the Reactor Coolant System With The Vacuum Skid Attached (Section E8.2).

50-400/99-02-01 NCV Failure to Comply With TS 3.3.2 (Section O1.2).

19 50-400/99-02-03 NCV Failure to Calibrate Instrumentation Required for Inservice Testing (Section M7.1).

50-400/99-02-04 NCV Inadequate Surveillance Procedures (5 Examples)

(Sections M7.1 and M8.2).

Discussed 50-400/1999-005-01 LER Engineered Safety Features Actuation Systems Technical Specifications exceeded (Section O8.2).