IR 05000261/1999004

Insp Rept 50-261/99-04 on 990523-0703.One Violation of NRC Requirements Occured & Being Treated as Ncv.Major Areas Inspected:Aspects of Licensee Operations,Maint,Engineering & Plant Support

ML14182A311
Person / Time
Site:	Robinson
Issue date:	07/30/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML14182A310	List: IR 05000261/1999004
References
50-261-99-04, NUDOCS 9908100112
Download: ML14182A311 (27)

Text

U. S. NUCLEAR REGULATORY COMMISSION

REGION II

Docket No:

50-261 License No:

DPR-23 Report No:

50-261/99-04 Licensee:

Carolina Power & Light (CP&L)

Facility:

H. B. Robinson Unit 2 Location:

3581 West Entrance Road Hartsville, SC 29550 Dates:

May 23 - July 3, 1999 Inspectors:

B. Desai, Senior Resident Inspector A. Hutto, Resident Inspector F. Jape, Senior Project Manager (Section E8.1)

J. Lenahan, Reactor Inspector (Section E7.1, E7.2)

C. Smith, Reactor Inspector (Section E7.1, E7.2)

W. Kleinsorge, Reactor Inspector (Section E7. 1, E7.2)

Approved by:

Brian R. Bonser, Chief Reactor Projects Branch 4 Division of Reactor Projects 9908100112 990730 PDR ADOCK 05000261 G

PDR Enclosure

EXECUTIVE SUMMARY H. B. Robinson Power Plant, Unit 2 NRC Integrated Inspection Report 50-261/99-03 This integrated inspection included aspects of licensee operations, maintenance, engineering, and plant support. The report covers a six-week period of resident inspection; in addition it includes the results of an engineering inspection and a year 2000 readiness program review by regional inspector Operations

The conduct of operations was professional, risk informed, and safety-conscious (Section 01.1).

• New fuel inspections were performed in accordance with the licensee's procedures. The required fuel inspections were performed and documented by a reactor enginee Foreign material exclusion procedures were correctly implemented (Section 01.2).

• Measurement of the end of life moderator temperature coefficient (MTC) was performed in accordance with licensee procedures. Plant maneuvers were conservative with respect to reactivity addition. There were no discrepancies noted with the acquisition of the data and the calculation of the MTC (Section 02.1).

• A system walkdown found that the residual heat removal system was appropriately configured and maintained. System parameters were being maintained within Technical Specification requirements (Section 02.2).

• A clearance associated with a service water system modification provided adequate isolation conditions for personnel safety and protection of plant equipment. The clearance was implemented in accordance with the licensee's procedures (Section 02.3).

Maintenance

Maintenance activities were conducted in accordance with applicable work documents and procedures. Personnel were properly trained and knowledgeable of their assignments (Section M 1.1).

• A review of completed surveillance test packages demonstrated acceptable test results (Section M2.1).

• _2 Engineering

The licensee's design change control procedures complied with the requirements of 10 CFR 50.59, and 10 CFR 50, Appendix B, Criterion Ill, and ANSI N45,2.11-1974 (Section E1.1).

• Plant modification packages were technically adequate with some exceptions. The 10 CFR 50.59 safety evaluation, design inputs, design evaluations, assumptions and references, and installation instructions generally met regulatory requirements. Six examples of a violation were identified for failure to follow design control procedures (Section E1.2).

• The licensee's procedure for inspection of the containment liner incorporated current regulatory requirements. Engineering provided good support to the plant to effectively implement industry experience with liner corrosion issues and prepare to implement the requirements of Article IWE of ASME Section XI (Section E2.1).

• Corrective actions taken to resolve safety injection accumulator level transmitter LT-930 inaccuracy involved numerous maintenance attempts over a two month period indicating poor problem resolution. While cross checking the operable accumulator level instrument LI-928, with other parameters showed that this instrument channel was functional, the licensee could have provided greater assurance of LI-928 accuracy by performing a calibration early in the corrective action process (Section 7.1).

• The self-assessments performed within the Robinson Engineering Support Section (RESS) were effective in identifying engineering performance deficiencies and were useful in providing oversight to management (Section 7.2).

• A year 2000 (Y2K) readiness program review was completed. Overall the Y2K project is about 98 percent complete and the contingency plan is about 95 percent complete. Both programs were on target to be completed by their scheduled due dates (Section 8.1).

Plant Support

Radiological controls and security practices were properly conducted. Areas observed in the radiological control area were appropriately posted and secured. The security plan was effectively implemented and compensatory actions were initiated when required (Section R1.1, S1.1).

• An instance where the licensee was going to use the motor driven fire pump for non-fire protection use was identified. Licensee procedures related to the fire protection system provided conflicting guidance with regard to this non-fire protection related use. The licensee plans to resolve this matter through the condition report system (Section F1.1).

Report Details Summary of Plant Status Robinson Unit 2 operated at 100 percent power throughout the inspection period with the exception that power was reduced on June 19 to 65 percent for approximately seven hours for turbine valve testin I. Operations

Conduct of Operations 0 General Comments (71707)

The inspectors conducted frequent control room tours to verify proper staffing, operator attentiveness and communications, and adherence to approved procedures. The inspectors routinely attended operations turnover meetings, management review meetings, and plan-of-the-day meetings to maintain awareness of overall plant operations. Operator logs, Condition Reports (CR), and instrumentation were routinely reviewed. Plant tours were conducted to verify operational safety and compliance with Technical Specifications (TS), as well as to assess plant housekeeping. In general, the inspectors concluded that the conduct of operations was risk informed, professional, and safety-consciou.2 New Fuel Assemblies Receipt Inspection (71707)

The inspectors observed receipt inspection by the licensee of two new fuel assembles per Fuel Handling Procedure FHP-002, "Unpacking and Handling of New Fuel Assemblies and Shipping Containers," Revision 10 and Fuel Management Procedure FMP-01 3, "New Fuel Assemblies and Shipping Containers Receipt Inspection,"

Revision The inspectors noted that maintenance and reactor engineering personnel involved had copies of the applicable procedures at the worksite and utilized them throughout the inspection. The inspectors observed security personnel perform the required security inspections of the new fuel containers as they were opened. The inspectors verified that the required inspection attributes were performed and documented in accordance with the procedures and that they were performed by a reactor engineer. The new fuel racks were controlled as a foreign material exclusion area (FMEA) as required per PLP-047,

"Foreign Material Exclusion Area Program," Revision 13, and personnel appropriately logged in to and out of the area. The inspectors also verified that the new fuel storage, locations were appropriately locked out for criticality control, as specified in FHP-00 No fuel assembly discrepancies were noted by the reactor engineer during the inspectio Operational Status of Facilities and Equipment 0 End Of Life (EOL) Moderator Temperature Coefficient (MTC) Measurement Inspection Scope (71707, 37551)

The inspectors observed operations personnel maneuver the plant to obtain data for determining EOL MTC per Engineering Surveillance Test, EST-146, "EOL MTC Measurement," Revision Observations and Findings Technical Specification (TS) SR 3.1.3.2 requires that the MTC be measured within seven days of the plant achieving the 300 ppm all rods out rated thermal power (ARO/RTP)

boron concentration. The purpose of the surveillance is to ensure that the MTC has stayed within the bounds of the design basis main steam line break analysis. The inspectors observed the performance of EST-146 to measure the MTC and satisfy the TS surveillance requiremen The inspectors observed the briefing for EST-146 and concluded that the briefing covered all the appropriate prerequisites, precautions and limitations. Particular emphasis was placed on coordination of control rod movements and turbine governor valve control where it was stressed that reactivity decreasing manipulations would always precede reactivity increasing manipulations. Reactor engineering personnel performed.the brief and were present during the test to collect the dat To perform the test, average reactor coolant system temperature was decreased degrees Fahrenheit. Data was collected before and after the temperature decreas Conditions were allow to stabilize prior to taking data. The inspectors verified that the data was accurately recorded. The inspectors also verified that axial flux limits as well as rod insertion limits were not exceeded during the tes The inspectors reviewed the completed test package and performed a check of the calculations for math errors. There were no errors noted. The MTC was determined by reactor engineering to meet the acceptance criteria of the test, thus meeting the TS requirement Conclusions Measurement of the end of life moderator temperature coefficient was performed in accordance with licensee procedures. Plant maneuvers were conservative with respect to reactivity addition. There were no discrepancies noted with the acquisition of the data and the calculation of the MT *

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0 Safety System Walkdown Inspection Scope (71707)

The inspectors conducted a general system walkdown of the residual heat removal (RHR) syste Observations and Findings The inspectors conducted a system walkdown of the RHR system to.assess the general condition of system components, including labeling, to verify that system valve positions matched the system drawings and station operating procedures, and to assess plant housekeeping and radiological conditions around system component No misaligned valves were identified. A minor discrepancy was noted with the system drawings which was pointed out to the system engineer for correction. The inspectors noted some boric acid buildup on the "A" RHR pump packing and on two drain cap These were pointed out to the system engineer who initiated a work request to have the components cleaned. No active leaks were noted. The inspectors also reviewed the applicable sections of the Updated Final Safety Analysis Report (UFSAR) and TS and identified no discrepancies. A review of the Maintenance Rule data base was also performed and the inspectors found that the appropriate performance criteria data were being collecte Conclusions A system walkdown found that the RHR system was appropriately configured and maintained. System parameters were being maintained within TS requirement.3 Clearance Walkdown (71707, 62707)

The inspectors verified proper implementation of clearance, 99-00629, during a walkdown on June 2. The clearance was to isolate the "B" service water (SW) booster pump header to allow tie-in of the SW cooling water injection modification ESR 97 00383. The inspectors verified that valves, electrical breakers, and control switches were aligned appropriately to provide an adequate boundary for the scheduled maintenance activity. No discrepancies were identified during verification of the clearance. The inspectors verified that the clearance was implemented in accordance with the licensee's procedure II. Maintenance M1 Conduct of Maintenance M Observation of Maintenance Activities (62707)

The inspectors observed all or portions of the following work requests (WR):

WR/JO 98-AEEI1, "Service Water Injection Cooling Modification Per ESR 97 00383"

WR/JO 99-ABXM1, "Repack "B" Service Water Booster Pump"

WR/JO AKHV 002, "Calibrate the Power Range Nuclear Instrument System Channel N-44" The inspectors determined that the maintenance observed was properly approved and was included on the plan of the day. The inspectors found that the work observed was thorough, and performed with the work package present and in use. Accompanying documents such as procedures and supplemental work instructions were properly followed. Personnel were properly trained and knowledgeable of their assignments. The inspectors noted that supervisors and system engineers monitored the jobs on a frequent basi M2 Maintenance and Material Condition of Facilities and Equipment M Review of Completed Surveillance Test Packages (61726)

The inspectors reviewed test package documentation for the following completed surveillance tests:

OST 402-2, "EDG "B" Diesel Fuel Oil System Flow Test," Revision 10

MST 021, "Reactor Protection Logic Train "B" At Power," Revision 20

OST 202, "Steam Driven Auxiliary Feedwater System Component Test," Revision

• EST 146, "EOL MTC Measurement," Revision 2 No problems were identified. Completed surveillance test packages demonstrated acceptable test result Ill. Engineering El Conduct of Engineering E Design Change Processes (37550) Inspection Scope The inspectors reviewed the licensee's procedures which control the design change progra Observations and Findinqs The inspectors reviewed the current revisions of the procedures listed below which control design and design changes to determine if the procedures implement the requirements of 10 CFR 50, Appendix B, Criterion Ill and 10 CFR 50.59. The following procedures were reviewed:

EGR-NGGC-0001, "Conduct of Engineering Operations" EGR-NGGC-0003, "Design Review Requirements EGR-NGGC-0005, "Engineering Service Requests" EGR-NGGC-0006, 'Vendor Manual Program" EGR-NGGC-0007, "Maintenance of Design Documents" EGR-NGGC-0012, "Equipment Database System Program" EGR-NGGC-0156, "Environmental Qualification of Electrical Equipment Important to Safety" EGR-NGGC-0320, "Civil/Structural Operability Reviews" EGR-NGGC-0351, "Performance Monitoring of Structures and Tanks" REG-NGGC-0002, "10 CFR 50.59 and Other Regulatory Evaluations" The inspectors verified that the procedures adequately addressed: design inputs, design calculations, drawing changes, post-modification testing, control of field changes, 10 CFR 50.59 safety evaluations, training, and as-low-as-reasonably achievable (ALARA) radiation dose review *

6 Conclusion The licensee's design change control procedures complied with the requirements of 10 CFR 50.59, and 10 CFR 50, Appendix B, Criterion III, and ANSI N45,2.11-197 E Design Changes and Plant Modifications Inspection Scope (37550)

The inspectors performed a review of modifications completed during the last refueling outage (RFO18), in progress, or planned for completion during the next refueling outage (RFO19) in order to verify compliance with regulatory requirements and the design control progra Observations and Findings The inspectors reviewed the 10CFR50.59 safety evaluation, design inputs, design evaluations, assumptions and references, installation instructions and drawings, and post modification testing instructions. The following documents the results of independent reviews performed to assess the technical adequacy of the plant modifications and their compliance with regulatory and ANSI N45.2.11-1974 requirement Engineering Service Request (ESR) 9800024, Station Battery B Cable Replacement, Revision 0 The objective of this plant modification was to replace the previously existing 4-1/c 500MCM cable which had a temperature rating of 60 degrees Celsius with a cable having a temperature rating of 90 degrees Celsius. A similar sized Rockbestos Firewall Ill cable with a temperature rating of 90 degrees Celsius and qualified to IEEE 383 was installed. The replacement cable was evaluated for ampacity, voltage drop and short circuit current capabilities. The higher temperature rating was determined to provide additional margin for cable ampacity based on station battery B loads. The calculation of record for evaluation of short circuit capabilities was based on 75 degrees Celsius. The higher temperature rating provided additional margin for short circuit protection. The voltage drop limitation was specified in order to ensure that station battery B sizing criteria was not impacted. Calculations RNP-E-6.023 and RNP-E-6.018 demonstrated that the 106.2 volts battery voltage criteria would be maintained if the installed cable length was less than 20 feet. The actual pull length for the cable was 16 feet which satisfied the voltage drop criteri The 10 CFR 50.59 screening evaluation completed for the plant modification was reviewed by the inspectors and determined to be acceptable. Calculations of record impacted by the plant modification were listed on the document update form (DUF) as a

follows:

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.Calculation RNP-E-6.004, DC Short Circuit Study

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Calculation RNP-E-6.005, Over Current Protection and Coordination of the 125 DC Distribution System (Trains A and B)

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Calculation RNP-E-6.018, DC Control Loop Study

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Calculation RNP-E-6.023, Minimum Inverter Voltage Verification Review of the current revision of calculation RNP-E-6.018, Revision 0, dated April 19, 1994, showed that the calculation was affected by 27 changes which had not been incorporated. Calculation RNP-E-6.023, Revision 3, dated December 16, 1997, was affected by six changes which had not been incorporated. The number of unincorporated calculation changes may have contributed to the use of incorrect design input for cable length in calculation RNP-E-6.031 prepared for ESR 9800345, discussed belo ESR 9800345, Station battery B Design Capacity, Revision 0 The objective of this plant modification was to replace the existing station battery B with a battery of a larger capacity. This modification will be implemented during RFO 19. The replacement battery will provide a battery discharge of 248 amperes for a period of one minute; a discharge of 125 amperes for an additional fifty eight minutes; and 157 amperes for a final one minute period while maintaining a minimum battery terminal voltage of 106.8 volts. The operation and function of station battery B motor control center (MCC) B and battery chargers B and B-1 will not be affected by the plant modification. The existing float voltage of 131 volts, plus or minus 1 volt, however, will be changed to 133 volts plus or minus 1 volt. The desired float voltage was specified as 132.6 volts DC with a maximum voltage of 134 volts DC and a minimum voltage of 132 volts DC. The replacement batteries will be procured from C&D Technologies, Inc. and will consists of a 125 volt, 60 cell battery, type KCR-1 1, complete with accessories including 2 spare cells. A battery rack for KCR-1 1 battery, 2 step model RD-803-13EP3, complete with hardware cell spacers, documentation and certifications will also be provided for the new installation The inspectors reviewed the results of the safety evaluation performed in accordance with the requirements of 10 CFR 50.59 for ESR 98-00345. This evaluation determined that a change to the FSAR was required. The FSAR describes the existing station Battery B manufacturer, model and capacity. The safety evaluation also concluded that the proposed plant modification did not involve an Unreviewed Safety Question (USQ).

The inspectors did not identify any deficiencies with the safety evaluatio The inspectors performed an independent design review of the plant modification package, including calculation RNP-E-6.031. This calculation provided the battery sizing and associated electrical calculations required to demonstrate the adequacy of the proposed replacement battery type. Revised inserts from the following calculations contained in calculation RNP-E-6.031 were reviewed by the inspectors:

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Calculation RNP-E-6.004, DC Short Circuit Study, Revision 3

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Calculation RNP-E-6.005, Over Current Protection and Coordination of the 125VDC Distribution System (Trains A and B), Revision 0

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Calculation RNP-E-6.020, Load Profile and Battery Sizing Calculation for Battery B, Revision 3

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Calculation RNP-E-6.022, DC Voltage Profile, Revision 3

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Calculation RNP-E-6.023, Minimum Inverter Verification, Revision 3

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Calculation RNP-E-6.024, Battery Charger Sizing, Revision 3

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Calculation RNP-E-5.018, Ampacity Evaluation of Safety Related Cables On 125VDC and 120 VAC MCCs and Buses, Revision 4

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Calculation RNP-E-9.002, Fuse Analysis for MCCs A and B, Battery Chargers A, B, A-1, and B-1, Revision 1 Based on review of the above calculations, the inspectors determined that the replacement batteries had been sized for a capacity of 127.58 ampere hours based on a one hour discharge to a terminal voltage of 1.81 volts DC per cell. Specifically, the cell sizing worksheet for cell type KCR-1 1, Figure 1 of Calculation RNP-E-6.020, demonstrated that the battery had been sized for an electrolyte temperature of 67 degrees Fahrenheit, with a design margin of 17 percent and an aging factor of 10 percent. The required cell size was determined to be 5 positive plates and was the basis for selecting cell type KCR-1 1. Calculation RNP-E-6.020 showed that the replacement battery had adequate capacity to meet its.enveloping load profile for 60 minutes based on the design margin and aging factor specifie The inspectors reviewed C&D Technology, Inc. vendor information and determined that the one hour capacity rating of the proposed battery was 204 ampere hours with an end of cell voltage of 1.75 volts. The vendor information also showed that a KCR/KAR-1 1 cell with a end voltage of 1.81 volts had a capacity of 320 ampere hours based on a four hour duty cycle. The one hour rating with an end voltage of 1.81 volts was listed as 173 ampere hours. The inspectors concluded that the proposed replacement battery "B" was adequately sized for a one hour duty cycle of 127.58 ampere hours with a 1.81 end of cell voltag The civil engineering evaluation of the plant modification determined that the new battery and battery racks were longer, heavier, and used a different support configuration and bolt layout. Calculation RNP-C/STRU-1 135, Station B Battery Rack and Anchorage, was prepared to address the new anchoring design. This calculation did not address the seismic qualification requirements of the battery and battery racks. It contained a note which alluded to the need for a owner's review of C&D Technologies, Inc. seismic calculations. Review of the civil engineering evaluation and the design basis section of the plant modification showed that applicable design input for specifying seismic requirements of the proposed battery and battery racks had not been included in the

design change package. Failure of the design change package to include the requirements of IEEE 344-1975 concerning seismic qualification of the new battery and battery racks was identified as the as a violation of 10 CFR 50, Appendix B, Criterion V, for failure to follow procedure EGR-NGGC-0005, Engineering Service Requests, Attachment 2, Design Inputs. This Severity Level IV violation is being treated as a Non Cited Violation (NOV), consistent with Appendix C of the NRC Enforcement Policy. This violation example is in the licensee's corrective action program as CR No. RNP 99 01217, Seismic Requirements of New Battery, dated June 21, 1999. It is identified as NCV 50-261/99-04-01, Failure to Follow Design Control Procedure Review of the electrical evaluation showed that calculation RNP-E-6-004, Short Circuit Study, failed to consider the as-installed length of battery B feeder cable replaced by plant modification ESR-98-0024 during RFO 18. The previously installed cable length of 15 feet was used in lieu of the pulled length of approximately 16 feet installed during RFO 18. Additionally, calculation RNP-E-6.031 specified a limit of 0.4 volts for the battery feeder cable voltage drop without documented evidence that this requirement would have been met with the presently installed cable length. The failure of the licensee to include applicable design input of cable length required to demonstrate compliance with maximum voltage drop limit and calculated short circuit current was identified as an additional example of a violation of 10 CFR 50 Appendix B, Criterion V, for failure to follow procedure EGR-NGGC-0005. This severity level IV violation is being treated as an NCV, consistent with Appendix C of the NRC Enforcement Policy. This violation example is in the licensee's corrective action program as CR No. RNP-99 01153, Reference to Cable Replacement, dated June 9, 1999. It is identified as NCV 50 261/99-04-0 ESR 9900182, 125 V DC Battery Chargers "A and B" Loading, Revision 0 Condition Report number 99-01155 documented a problem wherein station battery A and B was being loaded by their respective battery charger amplifier board and resistance capacitance (RC) filter. These chargers are equipped with a power failure disconnect relay (2CR) designed to strip the RC filter, amplifier board reference leg and magnetic amplifier from the charger DC output upon a loss of alternating current (AC)

power to the charger. The RC filter was determined to draw approximately 1 amperes from the battery when AC power is restored to the safety related MCCs supplying the chargers. The battery supplies this load until the charger magnetic starter is energized and the charger then supplies rectified AC power to the batter The objective of plant modification ESR 9900182 was to change the internal wiring configuration of the battery chargers to eliminate the RC filter from the circuit until the charger was energized. This ESR was initiated as corrective action for the problem identified in CR 99-01155. Additionally, the magnetic amplifier and amplifier board reference leg which previously was de-energized by the power failure disconnect relay would now be permanently wired into the circuit. This would ensure that voltage regulation had been established before energizing the magnetic contacto The 10 CFR 50.59 screening evaluation completed for the plant modification was reviewed by the inspectors and determined to be acceptable. The modified charger contribution to the battery load was identified as 0.5 amperes drawn by the magnetic amplifier and amplifier reference leg. Post modification test requirements were specified to measure the load after charger startup when the DC output breaker was close The following calculations of record and the plant modification were listed as requiring revision to incorporate the additional battery load:

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Calculation RNP-E-6.020, Load Profile and Battery Sizing Calculation for Battery B

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Calculation RNP-E-6.022, DC Voltage Profile

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Calculation RNP-E-6.023, Minimum Inverter Voltage Verification

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Calculation RNP-E-6.024, Battery Charger Sizing

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Calculation RNP-E-5.018, Ampacity Evaluation of Safety Related 125V DC and 120 V AC Power cables

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Calculation RNP-E-6.031, Station Battery B Replacement

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ESR 98-00345, Station Battery B Replacement The inspectors reviewed CP&L drawing No. 5379-3414, Three Phase Constant Potential Charger, and verified that the wiring changes shown by the sketches on page C2 and C3 of the plant modification package were correct and would achieve the design objectives specified. The inspectors did not identify any deficiencies with plant modification ESR 990018 ESR 9800509.North Service Water Header Replacement The inspectors reviewed ESR 9800509, Revision 6, North Service Water Header Replacement. This ESR reroutes the north service water header from its present buried route north around the radioactive waste building to an above ground route south of the radioactive waste building to the auxiliary building. Construction work was in progress on this modification during the inspectio The inspectors reviewed the 10CFR50.59 safety evaluation, design inputs, design evaluations, assumptions and references, installation instructions and drawings, and post modification testing instruction In addition to the above, the inspectors examined, for technical adequacy, the following:

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Sargent & Lundy Report, SL-7176, Service Water System Performance Test, dated July 18, 1889

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Specification 23869-C-203(Q), Receiving, Testing, Forming, Placing, Finishing and Curing of Concrete and Placing of Reinforcing Steel, Revision 0

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Specification 23869-C-31 1(Q), Purchase of Ready-mixed Concrete, Revision 0

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Calculation RNP-C/STRU-1 145, North Service Water Header Replacement Evaluation For Concrete Walls For Vehicle Barrier System Standoff Distance, Revision 0

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Calculation RNP-C/STRU-1 142, North Service Water Header Replacement Design of Concrete Walls

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Evaluation of Instrument Box Freeze Protection Heater Size (150W)

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PCI Energy Systems Welding Procedure Specification (WPS) 1 MN GTAW/SMAW, Revision 0, and supporting Procedure Qualification Record (PQR)

231 R/2 The inspectors conducted the following inspections: a walkdown of the modification installation area; observation of reinforcing steel installation; observation of concrete placement No. 4 activities including preparation, sampling, quality control batch testing, placement, consolidation and finishing activities; and observation of stud weld activitie The inspectors reviewed welding fabrication records for selected service water pipe replacement spool pieces, structural embedded plates and pipe supports. Records included: WPS; PQR; Welder or Welding Operator Performance Qualification (WPQ)

Certificates; Welders Continuity Logs; Welder Continuity Log History; welding filler material Certified Material Test Reports (CMTR)s; weld travelers; Nondestructive Examination (NDE) reports; NDE Level I & II personnel qualification certification documents; Vision Examination Reports; and Liquid Penetrant consumable materials CMTR The inspectors identified the following issues when reviewing the ESR and associated activities and documents:

Specification 23869-C-203(Q), Receiving, Testing, Forming, Placing, Finishing and Curing of Concrete and Placing of Reinforcing Steel, Revision 0, paragraph 4.2.2 states in part "Unless otherwise approved, spacers shall be factory-made wire bar supports...."

The inspectors noted that in some locations the reinforcing steel was supported, prior to concrete placement, by sections of "T' shaped structural steel in place of factory-made wire bar supports. The inspectors discussed the above with a licensee engineer who indicated that the substitution by sections of "T" shaped structural steel for factory-made wire bar supports had been verbally approved. The failure to document approvals for departures from specification requirements was identified as an additional example of violation 50-261/99-04-01, violation of 10 CFR 50, Appendix B, Criterion V for failure to follow procedure EGR-NGGC-0005. The licensee entered this issue into their corrective action program with CR No. 99-0122 The above ground route of the north service water header in ESR 9800509, Revision 6 necessitated the relocation of pressure transmitter PT-1616, North Service Water Header Pressure Transmitter and Service Water Low Pressure switches PSL-1616A &B from inside the auxiliary building to a spot outside. ESR 9800509 provided a heated enclosure for freeze protection for that instrumentation. ESR 9800509 did not however include a calculation to demonstrate the adequacy of the proposed freeze protection for the instrumentation tubing that connects those instruments to the service water pipe header. The failure to demonstrate the adequacy of the proposed freeze protection for the instrument tubing associated with pressure transmitter PT-1616 and pressure switches PSL-1616A & B in ESR 9800509 with a calculation was identified as another example of violation 50-261/99-04-01, violation of 10 CFR 50, Appendix B, Criterion V, for failure to follow procedure EGR-NGGC-0005. The licensee entered this issue into their corrective action program with CR No. 99-0123 ESR 9800509, Revision 6, Section C, paragraph 10.0, page C16 states in part "Welding shall be in accordance with AWS D1.1...Welded Stud Anchors shall be Nelson studs."

AWS D1.1-1988, Table 7.3.1 Mechanical Properties Requirements for Studs, specifies tensile strength, yield strength elongation and reduction of area requirements for welded stud anchors. The AWS D1.1, Table 7.3.1 mechanical property requirements were not referenced in the licensee's purchase order. As a consequence, the required certification of mechanical tests for the studs was not provided by the stud vendor. The licensee entered this issue into their corrective action program with CR No. 99-0903 This issue demonstrated a process interface breakdown between the design organization and the procurement organization. This was identified as another example of violation 50-261/99-04-01, violation of 10 CFR 50, Appendix B, Criterion V, for failure to follow procedure EGR-NGGC-000 The failure to document approval for a departure from specification requirements; failure to provide a calculation to demonstrate the adequacy of freeze protection; and an interface breakdown between the design and procurement organizations were identified as additional examples of NCV 50-261/99-04-0 ESR 9800072, ECCS Sump Debris Hood The inspectors reviewed ESR 9800072, Revision 0, dated December 22, 1998. This ESR provides the structural design details for installation of a cover (hood) over the ECCS sump. The.purpose of the hood is to prevent debris from falling vertically into the sump area, and thereby interfering with flow into the sump. The existing weir wall surrounding the sump prevents debris from being transported horizontally into the sum The inspectors reviewed the 1OCFR50.59 safety evaluation, design inputs, design evaluations, installation drawings and instructions, and marked-up copies of procedures affected by the modification. There are no post-modification testing requirements. The inspectors also reviewed calculation number RNP-C/STRUC-1 134, ECCS Sump Hood, Structural Design Analysis, Revision 0, dated December 16, 199 Review of the installation drawing, Number 9800072-C-1 000, ECCS Sump Hood, Elevation 228 Reactor Containment Building, Sheets 1 and 2 disclosed the following problems:

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The weld details for the connection of the tube steel columns to baseplates for members 2 & 3 were not shown on the drawing The weld details for the connection of the tube steel columns to baseplates for members 1 & 4 were not in accordance with the design calculation The licensee initiated CR 99-01254 to document and disposition these problem The above errors were indication of failure to perform design verification of design output documents. Failure to prepare accurate drawings was identified as an additional example of violation 50-261/99-04-01, violation of 10 CFR 50, Appendix B, Criterion V, for failure to follow procedure EGR-NGGC-0005. This violation example is in the licensee's corrective action program as CR No. 99-0125 In addition to the above weld symbol errors or deletions, the inspectors identified some administrative errors on the drawings including:

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Although a reference to Section G-G was shown on the drawings, the details for Section G-G were not included on the drawings. The licensee determined that Section G-G was not require Section H-H was mislabeled Section 1-A reference to a drawing note was not numbere A final calculation was not included in the ESR package for the as-designed baseplate expansion anchors. The design calculation included in the package was conservative, but did not reflect final design condition The above administrative issues were not included in the violation examples since they did not affect safety. However they were additional examples of lack of attention to detail ESR 9800427, Correct Instrument Tubing Slope The inspectors reviewed ESR 9800427. This ESR was issued as a portion of the corrective actions to resolve instrument slope deficiencies identified by NRC in Violation item 50-261/98-03-02. Implementation of this ESR will complete the licensee's corrective actions to closeout the instrument slope deficiencie The inspectors reviewed the 10 CFR 50.59 safety evaluation, the design inputs and design evaluations, and installation drawings and instructions. The drawings included Sketch numbers 9800427-C-1 000, Sheets 1 - 4, 9800427-C-1001, Sheets 1 - 4, and 9800427-C-1002, Sheet Conclusions Plant modification packages were technically adequate with some exceptions. The 10 CFR 50.59 safety evaluation, design inputs, design evaluations, assumptions and references, and installation instructions generally met regulatory requirements. Six examples of a non-cited violation (NCV) were identified for failure to follow design control procedure E2 Engineering Support of Facilities and Equipment E Inspection of Containment Vessel Liner Inspection Scope (37550)

The inspectors reviewed the licensee's program for inspection of the containment vessel liner which will be conducted during the next refueling outage (RFO 19). Observations and Findings The inspectors reviewed Special Procedure SP-1419, Inspection and Repair of CV Liner and Insulation, Revision 1, Effective Date June 1, 1999. This procedure establishes the inspection criteria, inspection methodology, and the documentation requirements for determining the condition of the containment liner during the next refueling outage (RFO 19). The procedure specifies inspection acceptance criteria and the requirements for engineering assessments and repair, if necessary. This procedure was updated to incorporate lessons learned from the inspections performed using Revision 0 of the procedure during RFO 18. The inspectors discussed the scope 'of the proposed inspections with licensee engineers, and the methodology to be used for selection of areas to be inspected. The inspectors also discussed the licensee's plans for implementation of-the requirements of 10 CFR 50.55a and Article IWE of ASME Section XI, which are effective September 9, 200 Conclusions The licensee's procedure for inspection of the containment liner incorporated current regulatory requirements. Engineering provided good support to the plant to effectively implement industry experience with liner corrosion issues and prepare to implement the requirements of Article IWE of ASME Section X E7 Quality Assurance in Engineering E Implementation of Corrective Action Program Inspection Scope (37550)

The inspectors reviewed corrective actions developed and implemented for various plant problems in order to verify compliance with the licensee's corrective action program and 10 CFR 50 Appendix B, Criterion XV Observations and Findings Corrective actions developed and implemented by the licensee for the following CRs were evaluated by the inspectors:

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CR No RNP 99-01155, Charger Impact on Battery Load dated June 9, 1999

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CR No. RNP-99-0091 1, "C" SI Accumulator Level Indicator LI-930 Out of Service, dated April 30, 1999

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CR No. RNP 99-01133, Level Transmitter LT-930 Out of Service, dated June 6, 1999 Corrective action plans developed for resolution of the plant problem documented on CR No. RNP 99-01155 was implemented by plant modification ESR 99-00182. The inspectors did not identify any deficiencies with this design change packag Work Request/ Job Order ( WR/JO) 99-ABS1 was initiated on April 24, 1999, to correct a plant problem when plant operators noted that the control room level indicators LI-930 and LI-928 differed by more than 6 percent. This exceeded the maximum allowed deviation of 5 percent specified in Operations Surveillance Test (OST) procedure OST 020, "Shiftly Surveillance," Revision 9. Review of this WR/JO revealed that trouble shooting activities were performed involving individual calibration checks of LI-930 and ERFIS computer data, both of which were within as found tolerance requirements. No corrective maintenance was performed and part 5 section 2 of the CR was checked to indicate that no further investigation was require On April 29, 1999, WR/JO 99-ABSY1 was initiated to correct the continuing problem with LI-930. The corresponding level transmitter, LT-930, was removed from service. The instrument lines were blown down and a response check was performed on LT-930 which operated correctly. LT-930 was returned to service and a channel check with LT 928 showed that LT-930 (LI-930) read 72.3 percent and LT-928 (LI-928) read 70.77 percent accumulator level respectively. Neither LT-930 nor LT-928 were calibrated prior to returning LT-930 to service in order to ensure that the both instrument channels were providing accurate readings of the accumulator borated water volum On April 30, 1999, CR No. 99-00911 was written to document the continuing problem with the level indicators. LI-928 and LI-930 differed by more than 6 percent, which exceeded the maximum allowed deviation of 5 percent specified in procedure OST-02 This CR was initiated because of the licensee's inability to perform a channel check of the accumulator level every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in accordance with Technical Requirements Manual section 4.2.1. This channel check is performed to demonstrate operability of the accumulator in accordance with the requirements of TS section 3.5.1.C. When a channel check cannot be performed comparison of the operable channel indication and status to other indication or status derived from independent instrument channels measuring the same parameter was acceptable for satisfying TS 3.5. On May 16, 1999, WR/JO 99-ABQ1 was initiated to document corrective maintenance performed for resolving the continuing problem with control room indicator LI-930 drifting out of specified tolerance. As found data for the existing level transmitter LT-930 was obtained after which it was replaced with a new transmitter. The as found data for the old transmitter was out of tolerance for every required input listed on the test data shee On May 20, 1999, the new transmitter LT-930 was calibrated. The as-left tolerances were within limits specified in calculation RNP-I-INST-1052, Accumulator Level Channel Accuracy and Scaling Calculation. The inspectors noted that the instruments had not been calibrated after maintenance activities were performed under WR/JO 99-ABS1 on April 24, 1999 and WR/JO 99-ABSY1 on April 29, 1999. Completion of this calibration activity provided reasonable assurance, via channel check, that both instrument channels were providing accurate and reliable readings of the accumulator borated water volume. Prior to May 20, 1999, the licensee had no calibration data to conclude that LI-928 was providing accurate reading On May 28, 1999, the operators determined that LI-930 was again drifting out of tolerance. WR/JO 99-ACBU1 was initiated to document corrective maintenance implemented for resolution of the continuing problem with LT-930. The instrument power supply was replaced and the transmitter low side instrument line was blown down. A calibration check of signal comparator, level controller LC-930, was performed and the level channel was returned to service on the same da On June 6, 1999, CR RNP-99-01133 was initiated to document that LT-930 was taken out of service because the instrument was drifting high in an erratic manner. The inspectors reviewed Part 1, Section 2 of the CR where potential operability and reportability concern was checked as not being applicable. In discussions with the licensee's engineering personnel the inspectors requested information concerning the basis for Part 1, Section 2 of the CR having been checked "no." The licensee did not have a documented basis for this conclusion and on June 24, 1999, the inspectors were given the following documents prepared by the licensee in response to this request:

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CP&L Memo from John P. Boska to Harold K. Chernoff, Subject: Accumulator C Level, dated June 23, 1999

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CAPS Closure Form( CAPS 99-01133) dated June 24, 1999

The CAPS Closure Form addressed inoperability of LI-930 (LT-930) and provided information which indicated accumulator C was operable despite the problems with LT 930. A description of the corrective maintenance activities performed to resolve LT-930 instrument inaccuracy was consistent with the WR/JO descriptions above. Information in support of accumulator C operability described the use of LT-928 and cross checking performed with this instrument against other parameters such as accumulator pressur Changes in accumulator level and pressure were readily observable to the operations staff on the plant computer trends (ERFIS) which were closely followed by LT-928 readings. The licensee concluded that these observations provided reasonable assurance that LT-928 was a valid indication of accumulator leve The inspectors agreed that SR 3.5.1.2 to verify accumulator C borated water volume can be satisfied with the use of LI-928. The prerequisite for using LI-928, however, was to verify that the instrument loop met instrument accuracy requirements delineated in calculation RNP-1-INST-1052 in accordance with the requirements of the licensee's instrument calibration program. Calibration activities were performed for the newly installed LT-930 on May 20, 1999. By cross channel check the licensee then verified that instrument loop LT-928 was within calibration tolerances and was providing valid and accurate readings of accumulator C borated water volume while repeated attempts were being made to correct LI-93 The inspectors concluded that information provided by the operations staff for corrective maintenance performed on April 26, 1999, showed poor problem recognition and untimely problem resolution. Trend plots for the ERFIS computer point revealed that LT 930 may have been degraded for longer than 210 days. Additionally, the extent of condition review of the plant problem performed by the instrument technician was limited in that it focused on individual calibration checks of LI-930 and the ERFIS computer point. CR No. 99-00911 was closed on the basis that no further investigation was required and the problem had been resolved. Poor maintenance practices were also demonstrated on May 5, 1999, by failure to perform calibration of LT-930 prior to returning it to service. On June 26, 1999, the licensee performed additional trouble shooting on instrument LT-930 and determined that the presence of debris in the instrument sensing line caused the erratic performance of this instrument. The debris was removed from the instrument sensing line and the instrument has performed satisfactory since that time. This indicated that the earlier replacement of LT-930 was unnecessar Conclusion Corrective actions taken to resolve safety injection accumulator level transmitter LT-930 inaccuracy involved numerous maintenance attempts over a two month period indicating poor problem resolution. While cross checking the.operable accumulator level instrument LI-928, with other parameters showed that this instrument channel was functional, the licensee could have provided greater assurance of LI-928 accuracy by

performing a calibration early in the corrective action proces E Quality Assurance Assessment and Oversight Inspection Scope The inspectors reviewed self-assessments performed within the Robinson Engineering Support Sectio Observations and Findings Self-assessments are part of the overall CP&L quality assurance program at Robinso The self-assessments are performed in accordance with CP&L procedure REG-NGGC 0003, Self-Assessment, Revision 0. The results of these assessments are categorized as strengths, or findings. Findings are classified as issues, weaknesses, or items for management consideration. The self-assessments reviewed by the inspectors were as follows:

Report No. 98-25, Training and Qualification Report No. 98-27, Environmental Qualification Assessment 99-42, Containment HVAC Assessment 99-43, Engineering Workload Management Assessment 99-44, Engineering Organizational Structure Two weaknesses and 13 items for management consideration were identified during Assessment 99-42. The inspectors reviewed CR No's. 99-00373, 99-00375, and 99-00376 which documented corrective actions to be performed to address the finding The inspectors verified that the corrective actions were adequate to address the finding The inspectors also verified that CRs were initiated to document and disposition findings from the other assessments as required by procedure REG-NGGC-000 Conclusions The self-assessments performed within the Robinson Engineering Support Section (RESS) were effective in identifying engineering performance deficiencies and were useful in providing oversight to managemen E8 Miscellaneous Engineering Issues E Year 2000 (Y2K) Readiness Program Review (TI 2515/141)

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 revie During the review, the licensee stated that the Y2K Readiness Project activities were 98 percent completed with contingency planning being about 95 percent complete, and that both programs were on target to be completed by their scheduled due date Conclusions regarding the Y2K readiness of the facility are not included in this repor 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, 199 E (Closed) Inspector Followup Item (IFI) 50-261/97-201-08: Evaluation of Transfer to Cold Leg Recirculation. This IFI concerned the acceptability of the calculated peak cladding temperature (PCT) during transfer to cold leg recirculation. The licensee provided additional information to NRC and committed to recalculate the PCT using revised computer software which was approved by NRC. The licensee will submit the reanalysis to NRC for review during the Fall of 1999. The licensee issued CR 99-01471 to track completion of the corrective actions to resolve this issu E (Closed) Violation Item 50-261/98-03-02: Inadequate Corrective Actions to Resolve Instrument Line Slope Deficiencies. The licensee responded to this violation in letter dated March 6, 1998. This violation concerned failure of the licensee to correct instrument slope deficiencies as stated in letters to NRC dated September 14, 1994 and February 21, 1996. Corrective actions to address this violation and resolve the instrument slope deficiencies included the following:

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Correction of the six slope deficiencies identified by NRC during the A/E inspection. (Inspection Report No. 50-261/97-201)

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Performance of walkdown inspections to identify instrument lines which do not meet slope requirements specified in CP&L procedure EGR-NGGC-0151, Evaluating Instrument Sensing Line Installation For instrument lines identified with slope deficiencies, performance of an evaluation to determine if, based on instrument performance history, additional corrective actions are necessar Completion of repairs to correct instrument slope deficiencies where necessary to address instrument performance problem The licensee submitted the results of the instrument tubing inspections to NRC in a letter dated July 30, 1998, Subject: Additional Information Concerning Instrument Tubing Slope Inspections and Result The inspectors reviewed the results of the licensee's walkdown inspections completed during RFO 18 which identified approximately 125 instruments with instrument lines which did not conform to the criteria specified in EGR-NGGC-0151. Licensee engineers performed an evaluation of the performance history for these instruments and determined that approximately 25 exhibited potential operational discrepancies which could be attributed to the instrument slope issues. The inspectors reviewed CR No' and 98-00606 and ESR numbers 9700257 and 9800427 which were issued to document and correct the deficiencies. The licensee corrected some of the slope discrepancies using existing maintenance procedures under the work request (WR/JO)

process. The inspectors reviewed the following WR/JOs issued to correct slope discrepancies: 98-AANG1, AANG2, AANH1 through AANH4, AANL1, AANL2, ABHB1, ABJX1, ABSL1, ACUHI1, ACQK1, AKJM1, AFJN1, AFJP1, AFJQ1, AFKY1, and AFLC Corrective actions to address this violation were also reviewed during the inspection documented in NRC IR number 50-261/98-02. All corrective actions to resolve this violation have been completed except those required under ESR 9800427. Completion of this ESR is being tracked as Action Item number 7 under CR 98-00545 in the licensee's corrective action program. Additional monitoring of instrument transmitters will also be performed during RFO 19 under Action Item number 6 of CR 98-0054 IV. Plant Support R1 Radiological Protection and Chemistry Controls R General Comments (71750)

The inspectors periodically toured the radiological control area (RCA) during the inspection period. Radiological control practices were observed and discussed with radiological control personnel including RCA entry and exit controls, survey postings, locked high radiation area controls, and radiological area material condition. The inspectors concluded that radiation control practices were being conducted in accordance with procedures. The inspectors also toured the radwaste building and found that radwaste storage containers and laundry bags were in good condition and appropriately labeled. In addition, outside radwaste storage areas and structures were properly posted and exhibited correct labeling and effective housekeeping. The inspectors found that housekeeping throughout the plant was effective in maintaining areas free of unnecessary equipment and debris. Relatively few contaminated areas were noted, and posted locked high radiation areas were properly secured against unauthorized entr SI Conduct of Security and Safeguards Activities S General Comments (71750)

During the period, the inspectors toured the protected area and noted that the perimeter fence was intact and not compromised by erosion or disrepair. Isolation zones were maintained on both sides of the barrier and were free of objects which could shield or conceal an individual. Lighting of the perimeter and of the protected area was acceptabl F1 Control of Fire Protection Activities F Use of Fire Water Inspection Scope (71750)

The inspectors reviewed the status of the fire protection systems and questioned an instance where the licensee was going to use the Unit 2 motor driven fire pump (MDFP)

to water grass onsit Observations and Findings The inspectors reviewed procedure OMM-002, "Fire Protection Manual," Revision 2 This procedure stated that "fire protection equipment is only for the use for emergencies, drill, or training." A contradictory statement in the same procedure stated "the use of fire water for cleaning and for the required purposes was allowed following approval of the Superintendent of Shift Operations." Additionally, National Fire Protection Association NFPA-24, "Installations of Private Fire Service Mains and Their Appurtenances," states that "usage of fire hydrants and fire hoses for non-fire service activities was not permitted. UFSAR Section 9.5.1 references NFPA as being used for the design criteria of fire water system The inspectors discussed the issue with the responsible fire protection engineer as well as the fire protection technician. Any further use of fire protection equipment was stopped pending further review. A condition report, CR-99-01247 was initiated. The inspectors were informed by the licensee that the planned use of the fire pump for watering grass was an isolated cas Conclusions An instance where the licensee was going to use the MDFP for non fire protection use was identified. Licensee procedures related to the fire protection system provided conflicting guidance with regard to this non fire protection related use. The licensee plans to resolve this matter through the condition report syste 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 July 12, 1999. The licensee acknowledged the findings presented at the exit meeting. Dissenting comments were not received from the licensee. The licensee did not identify any materials used during the inspection as proprietary information. No proprietary information was identifie PARTIAL LIST OF PERSONS CONTACTED Licensee J. Boska, Superintendent, Electrical Engineering, Robinson Engineering Support Section (RESS)

T. Cleary, Manager, Operations H. Chernoff, Supervisor, Licensing/Regulatory Programs J. Clements, Manager, Site Support Services R. Duncan, Manager, Robinson Engineering Support Services W. Farmer, Supervisor, ECCS, RESS J. Fletcher, Manager, Maintenance J. Moyer, Director, Site Operations R. Steele, Manager, Outage Management D. Stoddard, Superintendent, Mechanical Engineering, RESS T. Walt, Plant General Manager R. Warden, Manager, Regulatory Affairs A. Williams, Manager, Training P. Yandow, Licensing Engineer D. Young, Vice President, Robinson Nuclear Plant

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NRC B. Desai, Senior Resident Inspector A. Hutto, Resident Inspector INSPECTION PROCEDURES USED IP 37550:

Engineering IP 37551:

Onsite Engineering IP 61726:

Surveillance Observations IP 62707:

Maintenance Observation IP 71707:

Plant Operations IP 71750:

Plant Support Activities IP 92903:

Followup - Engineering TI 2515/141 Review of Year 2000 (Y2K) Readiness of Computer Systems at Nuclear Power Plants IP 92903:

Followup-Engineering

ITEMS OPENED AND CLOSED Opened 50-261/99-04-01 NCV Failure to Follow Design Control Procedures (Section E1.2)

Closed 50-261/99-04-01 NCV Failure to Follow Design Control Procedures (Section E1.2).

• 50-261/97-201-08 IFI Evaluation of Transfer to Cold Leg Recirculation (Section E8.2).

50-261/98-03-02 VIO Inadequate Corrective Actions to Resolve Instrument Line Slope Deficiencies (Section E8.3).