IR 05000261/2007006

IR 05000261-07-006; on 07/16/2007 - 07/20/2007, 07/30/2007 - 08/03/2007, 08/13/2007 - 08/16/2007; H.B. Robinson Nuclear Plant; Component Design Bases Inspection

ML072681194
Person / Time
Site:	Robinson
Issue date:	09/21/2007
From:	Binoy Desai NRC/RGN-II/DRS/EB1
To:	Walt T Carolina Power & Light Co
References
IR-07-006
Download: ML072681194 (34)

Text

ber 21, 2007

SUBJECT:

H. B. ROBINSON NUCLEAR PLANT - NRC COMPONENT DESIGN BASES INSPECTION REPORT 05000261/2007006

Dear Mr. Walt:

On August 16, 2007, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at your H. B. Robinson Nuclear Plant. The enclosed inspection report documents the inspection findings which were discussed on August 16, 2007, with you and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

Based on the results of this inspection, the inspectors identified one finding of very low safety significance (Green). This finding was determined to not involve a violation of NRC requirements. During this inspection your staff identified one finding of very low safety significance (Green). This finding was determined to involve a violation of NRC requirements.

However, because of the very low safety significance and because it was entered into your corrective action program, the NRC is treating the finding as a non-cited violation consistent with Section VI.A.1 of the NRCs Enforcement Policy. If you deny this finding or non-cited violation you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the United States Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington DC 20555-0001, with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the H. B.

Robinson Nuclear Plant.

CP&L 2 In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Binoy Desai, Chief Engineering Branch 1 Division of Reactor Safety Docket No.: 50-261 License No.: DPR-23

Enclosure:

NRC Inspection Report 05000261/2007006 w/Attachment:

Supplemental Information

REGION II==

Docket No.: 50-261 License No.: DPR-23 Report No.: 05000261/2007006 Licensee: Carolina Power & Light Facility: H. B. Robinson Nuclear Plant, Unit 2 Location: Hartsville, SC 29550 Dates: July 16 - August 16, 2007 Inspectors: S. Rose, Senior Reactor Inspector (Lead)

D. Mas-Penaranda, Reactor Inspector C. Peabody, Reactor Inspector M. Speck, Resident Inspector H. Anderson, Contractor G. Nicely, Contractor Approved by: Binoy Desai, Chief, Engineering Branch 1 Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS

IR05000261/2007006; 7/16/2007 - 7/20/2007, 7/30/2007 - 8/3/2007, 8/13/2007 - 8/16/2007; H.

B. Robinson Nuclear Plant; Component Design Bases Inspection.

This inspection was conducted by a team of four NRC inspectors and two NRC contractors.

Two green findings were identified during this inspection. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using IMC 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649,

Reactor Oversight Process, Revision 3, dated July 2000.

Cornerstone: Mitigating Systems

Green.

The team identified a finding having very low safety significance (Green)involving the failure of the licensee to meet a self imposed standard. The licensee committed in modification package EC 59037, Install D Deep Well Pump, to meet or exceed the requirements in the Electrical Power Distribution System Design Basis Document (DBD), DBD/R87038/SD16. DBD sections 4.3.1.c and 4.5.1.20 specified that overload protection be provided. The vendor technical manual for the D deep well pump motor, which is included in the facility technical manual 762-209-103 for the D deep well pump, specified that Thermal Overload (TOL) protection be provided. The vendor technical manual for the D deep well pump motor was referenced in modification package, EC 59037. Contrary to the above, the licensee failed to install TOL protection for the D deep well pump.

This finding was more than minor based on the fact that it is associated with the reactor safety mitigation cornerstone aspect of design control. It impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, the team conducted a Phase 1 SDP screening and determined the finding was of very low safety significance (Green). Since the D deep well pump is not safety related equipment per Chapter 15 of the UFSAR, this finding does not represent a violation of any NRC requirements. The team did not identify any cross cutting aspects associated with this finding. This issue is documented in the corrective action program as nuclear condition report (NCR) 239915. (Section 1R21.2.9.)

B. Licensee-identified Violations One violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meet the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a Non-Cited Violation.

The violation and corrective action is listed in Section 4OA7 of this report.

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Mitigating Systems and Barrier Integrity

1R21 Component Design Bases Inspection

.1 Inspection Sample Selection Process

The team selected risk significant components and operator actions for review using information contained in the licensees Probabilistic Risk Assessment (PRA). In general, this included components and operator actions that had a risk achievement worth factor greater than two or Birnbaum value greater than 1 X10-6. The components selected were located within the service water cooling (SW) system, chemical volume and control (CVCS) system, emergency diesel generator (EDG) electrical subsystems, station battery system, boric acid system, auxiliary feedwater (AFW) system, and component cooling water (CCW) system. The sample selection included 20 components, 6 operator actions, and 6 operating experience items. Additionally, the team reviewed two modifications by performing activities identified in IP 71111.17, Permanent Plant Modifications, Section 02.02.a. and IP 71111.02, Evaluations of Changes, Tests, or Experiments.

The team performed a margin assessment and detailed review of the selected risk-significant components to verify that the design bases have been correctly implemented and maintained. This design margin assessment considered original design issues, margin reductions due to modification, or margin reductions identified as a result of material condition issues. Equipment reliability issues were also considered in the selection of components for detailed review. These included items such as failed performance test results, significant corrective action, repeated maintenance, maintenance rule (a)1 status, RIS 05-020 (formerly GL 91-18) conditions, NRC resident inspector input of problem equipment, system health reports, industry operating experience and licensee problem equipment lists. Consideration was also given to the uniqueness and complexity of the design, operating experience, and the available defense in depth margins. An overall summary of the reviews performed and the specific inspection findings identified is included in the following sections of the report.

.2.1 CCW Pumps

a. Inspection Scope

Electrical: The team identified the design basis functions of the component and verified by review of schematic drawings, that operation of the pump motors ware consistent with the design basis and operational requirements. The team reviewed the protection coordination calculation for the CCW pump motor and verified that the circuit breaker ratings and protective devices trip settings and alarm functions were consistent with the licensing basis and operational requirements. The team verified that the Brake Horsepower (BHP) required by the pump is within the motor rating. The team reviewed the AC voltage calculations to ensure satisfactory voltage existed to the motors under worst case conditions. The team performed a walkdown of the CCW system to assess observable material condition of the pump motors. The team verified that the ambient conditions were consistent with vendor recommendation for the motors.

Mechanical: The team reviewed the CCW DBDs, Updated Final Safety Analysis Report (UFSAR), and applicable plant drawings to identify the design bases requirements of the equipment. The team examined the machinery history of the Component Cooling Water (CCW) pumps to verify that design bases have been maintained. The team examined records and test data for both corrective and preventative maintenance, and applicable corrective actions to verify that potential degradation was being monitored, prevented, and corrected. The team verified the preventative maintenance history and schedule is consistent with vendor recommendations. The team also reviewed vibrational testing and lube-oil testing for the CCW Pumps to verify the likelihood of pump damage or failure from these mechanisms. The team reviewed the Net Positive Suction Head (NPSH) calculation to verify that the CCW pumps were capable of performing their design function during accident conditions. The team also conducted a field walkdown of the CCW pumps with the CCW System Engineer to verify that the installed configuration is consistent with the design basis and plant drawings.

b. Findings

No findings of significance were identified.

.2.2 CCW Relief Valve CC-729

a. Inspection Scope

The team reviewed the CCW DBDs, UFSAR, and applicable plant drawings to identify the design bases requirements of the equipment. The team examined the machinery history of the CCW Relief Valve, CC-729 to verify that design bases had been maintained. The team examined records and test data for both corrective and preventative maintenance to verify that potential degradation was being monitored, prevented, and corrected. The team also verified that the maintenance, test, and inspections were being conducted in accordance with vendor recommendations and ASME code requirements.

b. Findings

No findings of significance were identified.

.2.3 CCW Surge Tank Level Instrumentation

a. Inspection Scope

The team reviewed DBDs, UFSAR, and applicable plant drawings to verify that the design bases and design assumptions have been appropriately translated into design calculations and procedures. The team reviewed the System Health and Status Report to assess the system equipment performance, reliability, maintenance, and material indicators/issues. The team reviewed the instrumentation setpoints, supporting calculations/bases, surveillance/calibration procedures, and calibration intervals for the level transmitters to ensure technical adequacy. The team reviewed the vendor instruction manual for the level transmitters to verify the capability to perform their function. The team verified that instrumentation and alarms were available to operators for making necessary decisions. The team additionally reviewed the automatic action of Flow Control Valve (FCV) FCV-626, which is required to shut on high flow rate for a ruptured tube in a reactor coolant pump thermal barrier cooling coil. The operating voltage for FCV-626 and its control circuit were reviewed to ensure proper operation during normal and degraded voltage conditions which would prevent overflow of the CCW surge tank.

b. Findings

No findings of significance were identified.

.2.4 CCW Piping (Integrity)

a. Inspection Scope

The team reviewed the CCW DBD, UFSAR, and applicable plant drawings to identify the design basis requirement of the equipment. The team reviewed the system history of the Auxiliary Building CCW piping to verify that the design basis has been implemented and maintained. The team examined records and test data for both corrective and preventative maintenance, and applicable corrective actions to verify that potential degradation was being monitored, prevented, and corrected. The team compared the Preventative Maintenance (PM) history and schedule with ASME requirements. The team also conducted a field walkdown of the system piping with the respective system engineers to verify that the installed configuration is consistent with the design basis and plant drawings.

b. Findings

No findings of significance were identified.

.2.5 Auxiliary Building SW Piping (Integrity and Flooding)

a. Inspection Scope

The team reviewed the SW and EDG DBDs, UFSAR, and applicable plant drawings to identify the design basis requirement of the equipment. The team reviewed the system history of the Auxiliary Building SW piping to verify that the design basis has been implemented and maintained. The team examined records and test data for both corrective and preventative maintenance, and applicable corrective actions to verify that potential degradation was being identified, monitored, and addressed. The team compared the PM history and schedule with ASME requirements. The team also conducted a field walkdown of the system piping and EDG floor drains with the respective system engineers to verify that the installed configuration is consistent with the design basis and plant drawings.

b. Findings

No findings of significance were identified.

.2.6 Manual Valves SW-118 and AFW-24 (backup to supply AFW)

a. Inspection Scope

The team reviewed the SW and AFW DBDs, UFSAR, and applicable plant drawings to identify the design bases requirements of the equipment. The team examined the testing, inspection, and machinery history of manual valves SW-118, AFW-24, and the associated piping to verify that design bases have been maintained. The team reviewed plant test procedures and results to verify that established acceptance criteria were met.

The team examined records and test data for both corrective and preventative maintenance to verify that potential degradation was being monitored and/or prevented.

The team also conducted a field walkdown of these valves with the respective system engineers to verify that the installed configuration was consistent with the design basis and plant drawings.

b. Findings

No findings of significance were identified.

.2.7 Steam Driven Auxiliary Feedwater (SDAFW) Pump

a. Inspection Scope

The team reviewed the AFW DBD, UFSAR, and applicable plant calculations, procedures, and drawings to identify the design bases requirements of the equipment.

The team examined representative machinery history related to the SDAFW pump to verify that design bases associated with the pump have been maintained. The team examined records and test data for both corrective and preventative maintenance as well as periodic surveillance testing, and applicable corrective actions to verify that potential degradation was being monitored and prevented or corrected. The team reviewed the availability of water from the suction source originating from the Condensate Storage Tank (CST), as well as backup sources from the SW system and D Deep Well pump. These reviews included available water volumes and levels, associated instrumentation, provision of adequate NPSH, and vortex protection to the SDAFW pump to verify that the SDAFW pump was capable of performing its design function during accident conditions. The reviews included verification of the adequacy of the provisions to ensure adequate minimum flow protection for the pump. The reviews also included verification of related aspects of hydraulic models and calculations developed to demonstrate the capability of the SDAFW pump and the AFW system to provide system flows and developed heads in accordance with design basis capabilities stated in the UFSAR and other design basis documents. The team reviewed the licensees establishment, review, and maintenance of pump performance and test criteria. This review demonstrated how the in-service testing program verified that ASME Code requirements, as well the minimum safety analysis design capabilities, were verified during routine surveillance testing and during post maintenance and modification testing. The team also conducted a field walkdown with the AFW system engineer of the SDAFW pump and associated support equipment and systems to verify, by visual observation of reasonably accessible locations, that the installed configuration and material condition of the pump were consistent with the design basis and plant drawings.

b. Findings

No findings of significance were identified.

.2.8 AFW Pump Discharge Check Valves (AFW-40, 41, and 84)

a. Inspection Scope

The team reviewed the AFW DBDs, UFSAR, and applicable plant drawings to identify the design bases requirements of the equipment. The team examined the machinery history of AFW Pump discharge check valves AFW-40, AFW-41, and AFW-84 to verify that design bases have been maintained. The team examined records and test data for both corrective and preventative maintenance, as well as reviewed applicable corrective actions to verify that potential degradation was being monitored and/or prevented and that these programs were in accordance with ASME code requirements and vendor recommendations. The team also conducted a field walkdown of these check valves with the AFW system engineer to verify that the installed configuration was consistent with the design basis and plant drawings.

b. Findings

No findings of significance were identified.

.2.9 D Deep Well Pump

a. Inspection Scope

(Closed) URI 05000261/2006005-001, Equipment Performance for Functional Recovery After Certain UHS Scenarios. During previous inspections (05000261/2006004 and 05000261/2006005), inspectors identified an issue related to the ultimate heat sink configuration and operational controls. This issue required further NRC review to determine whether the installed equipment could support recovery of SW in certain accident scenarios. The following was conducted during this component design basis inspection to close this URI:

Electrical: The team reviewed the SW design basis document, system description, and applicable UFSAR sections describing the function and basis for the D deep well pump. The team verified that design bases and design assumptions have been appropriately translated into design calculations and procedures. The team reviewed the applicable operating procedures for the D deep well pump for adequacy. The team verified that instrumentation and alarms were available to operators for making necessary decisions. The team reviewed the normal and alternate power supply sources and manual transfer schemes for the D Deep Well Pump motor, including those used for control functions, to verify that they would be available and have capacity and capability during the required design bases conditions. The team reviewed the protection/coordination of the electrical equipment to identify if the D deep well pump motor and cable was protected with properly set protective devices, identify if selective coordination existed with upstream protective devices, and identify if spurious operation would occur during operation of the motor for normal and abnormal events. The team reviewed calculations to verify that the feeder breakers were within design capabilities under maximum calculated fault conditions. The team reviewed calculations to verify that analyses validated component operation under required operating conditions (i.e.

confirm that the design basis minimum voltage at the motor terminals and controls would be adequate for starting and running the motor during required design bases).

The team additionally reviewed modification EC 59037, Install D Deep Well pump to determine if the design bases and design assumptions were appropriately translated into an adequate design. The team performed a visual non-intrusive inspection of the D deep well pump installation, including electrical transfer switches, cable installations, and installed connections to the emergency diesel generator heat exchangers to assess the installation configuration, material condition, and potential vulnerability to hazards.

Mechanical: The team examined available initial testing, design, and periodic performance testing documentation to verify that design bases associated with the D Deep Well pump have been developed, documented, and maintained. The team reviewed the available UFSAR information and applicable plant calculations, procedures, and drawings related to the overall integrated purpose and function of the D Deep Well pump to verify that the design bases are appropriately established and maintained. The team examined records and data for the annual performance inspection of the pump which have been recorded since the pump was placed in service to verify that potential degradation was being monitored and prevented or corrected.

The team reviewed the availability of water from the suction source originating from the D Deep Well, well draw down test results, and documentation of provision of adequate net positive suction head and minimum submergence levels to support pumping requirements in events where the pump would be required. The review included verification of the adequacy of the provisions to ensure adequate minimum flow protection for the pump. The review also included verification of related aspects of the associated thermal-hydraulic model and the End Path Procedure (EPP) and EPP Bases document to demonstrate the capability of the pump and the associated modification features, as well as the interface with then existing systems, to provide system flows, developed heads, required heat transfer capability, and runout protection in accordance with design basis capabilities and requirements stated in the design basis documentation. The review also included the calculations and assumptions related to determination of time available, in the event of loss of the Robinson Dam Tainter Gates, until the ultimate heat sink would be considered lost. This included the licensees informal estimation of the available Ultimate Heat Sink (UHS) volume which could be drawn from the Robinson Lake. The review included assessment of the thermal-hydraulic calculation, in regard to its determination of times available in the event the intake structure were lost until subsequent distinct procedural actions would be required, in comparison to the time requirements as stipulated in the EPP and EPP Bases document to confirm consistency in the time-step requirements. The team also conducted a field walkdown with the system engineer of plant features associated with the D Deep Well pump and its design purpose to verify, by visual observation of reasonably accessible locations, that the installed configuration and material condition of the features are consistent with the design basis and plant drawings.

b. Findings

As a result of this inspection, URI 05000261/2006005-001, Equipment Performance for Functional Recovery After Certain UHS Scenarios, was closed. However, a finding was identified related to the D deep well pump.

Introduction.

The team identified a finding having very low safety significance (Green) involving the failure of the licensee to meet a self imposed standard. The licensee committed in the Engineering Change (EC) package EC 59037, Install Deep Well Pump D, to meet or exceed the requirements in the DBD Electrical Power Distribution System, DBD/R87038/SD16. DBD sections 4.3.1.c and 4.5.1.20 specify that overload protection be provided. Vendor technical manual 762-209-103 for the D deep well pump specified that TOL protection is required. The vendor technical manual for the D deep well pump motor was referenced in the modification package EC 59037.

Description.

The performance deficiency was that the licensee failed to provide thermal overload relay protection in the design of the D Deep Well pump motor implemented under modification package EC 59037, Deep Well Pump D, although required by vendor documentation and DBDs.

Specifically, the licensee installed the D Deep Well pump as a part of modification package EC 59037. The licensee committed in EC 59037 to meet or exceed the requirements in the DBD Electrical Power Distribution System, DBD/R87038/SD16 to specifically meet the requirements of ANSI N45.2.11, Quality Assurance Requirements for the Design of Nuclear Power Plants. EC 59037 section B.4.13 states: Design Basis Document DBD/87038/SD16 defines the functional requirements, regulatory requirements, commitments relative to system design, and the original design codes and standards of record for the electrical distribution system. The design of the EC is within the boundary and scope of these Design Basis Documents, and to the extent defined below will meet or exceed the requirements of the documents through the utilization of specific design standards, guides, codes, and statements identified herein.

DBD Section 4.3.1.c states: The emergency on-site ac power system supports the cable and raceway system in that it provides overload and short circuit protection for power distribution and load feeder cables. DBD section 4.5.1.20 states: The on-site emergency ac power system shall be provided with protective devices for overload and short circuit protection. The vendor technical manual from Grundfos and Franklin Electric for the D deep well pump specified that each of the three motor legs must be protected with ambient-compensated extra quick-trip thermal overload relays. The vendor technical manual for the D deep well pump motor was referenced in the modification package EC 59037. The vendor requirements are due to the characteristics of submersible motors being different from standard motors and special overload protection is required. Thermal overload relays prevent an electric motor from drawing too much current and overheating. Thermal overload conditions are the most likely faults to be encountered in industrial motor applications and result in a rise in the motor running current, which produces an increase in the motor's thermal dissipation and temperature. Overload protection prevents an electric motor from drawing too much current, overheating, and failing due to the motor windings burning out. During periodic testing of the D Deep Well pump, failure to have an adequate protection scheme in place, places the pump motor at risk for damage or failure when required to perform its design function. During periodic testing, should the pump or motor become degraded or become overloaded, the motor could become damaged to the point of having an undetectable failure the next time the motor is required to be operated.

Having properly sized thermal overload relay protection installed would prevent this from occurring.

Analysis.

The team determined that the issue was a performance deficiency that was within the licensees ability to foresee and correct, and that is could have been prevented upon adequate review of the vendor submitted technical manual. The team determined that this finding was more than minor because it is associated with the reactor safety mitigation cornerstone aspect of design control and it impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, the team conducted a Phase 1 SDP screening and determined the finding was of very low safety significance (Green) because the absence of TOL protection did not result in the loss of operability of the D deep well pump. This issue is documented in the corrective action program as NCR 239915. Since the D deep well pump is not safety related equipment per Chapter 15 of the UFSAR, this finding does not represent a violation of any NRC requirements. The team did not identify any cross cutting aspects associated with this finding. (FIN 05000261/2007006-01, Failure to install Thermal Overload (TOL)protection on the D deep well pump.)

Enforcement.

No violation of regulatory requirements occurred.

.2.10 Alternate Cooling Sources to the Charging Pumps

a. Inspection Scope

The team reviewed the capability of the alternate cooling sources to remove heat from the charging pump seals. The team reviewed the CVCS operations training system description and compared the capacity of the alternate sources to the primary source of fluid drive oil cooling to verify that the proposed alternative water sources would be able to maintain the system cooling requirements.

b. Findings

No findings of significance were identified.

.2.11 Seal Injection Filter

a. Inspection Scope

The team examined representative corrective and preventative maintenance and procedural surveillance histories related to the seal injection filters, the filter high Differential Pressure (DP) alarm, and alarm setpoint, as well as associated design bases related to the filter capacities as described in the CVCS description and DBD, UFSAR, and applicable plant calculations, procedures, drawings, and the vendor manual to confirm that the design bases are appropriately implemented and maintained in relation to the operation and maintenance of the filter system. The team examined records of corrective and preventative maintenance and applicable corrective actions to verify that potential degradation was being monitored and prevented or corrected and that filter change outs were conducted at criteria consistent with the filter system design.

The team verified the filter DP instrumentation was appropriately maintained and monitored in regard to both indication and high alarm. The team also conducted a field walkdown with the system engineer of plant features associated with the seal injection filters and their design purpose to verify, by visual observation of reasonably accessible locations, that the installed configuration and material condition of the features are consistent with the design basis and plant drawings.

b. Findings

No findings of significance were identified.

.2.12 CVCS Pneumatic Flow Control Valves FCV-113B and FCV-114B

a. Inspection Scope

The team reviewed the CVCS DBD and applicable plant drawings to identify the design bases requirements of the equipment. The team examined machinery history of the CVCS pneumatic flow control valves FCV-113B and FCV-114B to verify that design bases have been maintained. The team reviewed plant test procedures and results to verify that established acceptance criteria were met. The team examined records and test data for both corrective and preventative maintenance, as well as reviewed applicable corrective actions to verify that potential degradation was being monitored and/or prevented. The team also conducted a field walkdown of these valves with the CVCS system engineer to verify that the installed configuration is consistent with the design basis and plant drawings.

b. Findings

No findings of significance were identified.

.2.13 Boric Acid Transfer Pumps

a. Inspection Scope

The team reviewed the CVCS description and DBD, UFSAR, and applicable plant calculations, procedures, and drawings, to identify the design bases requirements of the equipment. The team examined representative machinery history related to the Boric Acid Transfer (BAT) pumps to verify that design bases associated with the pumps have been maintained. The team examined records and test data for both corrective and preventative maintenance as well as periodic surveillance testing, and applicable corrective actions to verify that potential degradation was being monitored and prevented or corrected. The team reviewed the availability of the suction source to the pumps originating from the boric acid tanks. These reviews included available water volumes, levels, associated instrumentation, and pump suction configuration from the boric acid tanks to verify the conclusions of the existing calculation of available NPSH.

Although there was no existing formal calculation of vortex protection, the team reviewed and concurred with the licensees assessment of vortex protection which was developed during the course of the inspection. The reviews included verification of the adequacy of the provisions to ensure adequate minimum flow protection for the pump.

The reviews also included verification of related aspects of calculations demonstrating the capability of the BAT pumps to provide system flows and developed heads in accordance with design basis capabilities stated in the UFSAR and other design basis documents. The team reviewed the licensees establishment, review, and maintenance of pump performance and test criteria. This review demonstrated how the in-service testing program verified that ASME Code requirements, as well the minimum safety analysis design capabilities, were verified during routine surveillance testing and during post maintenance and modification testing. The team also conducted a field walkdown with the CVCS system engineer of the BAT pumps and associated support equipment and systems to verify, by visual observation of reasonably accessible locations, that the installed configuration and material condition of the pumps are consistent with the design basis and plant drawings.

b. Findings

No findings of significance were identified.

.2.14 Boric Acid Filter

a. Inspection Scope

The team reviewed the CVCS description and DBD, as well as associated design bases related to the filter capacities, UFSAR, and applicable plant calculations, procedures, drawings, and the vendor manual to identify the design bases requirements of the equipment. The team reviewed representative corrective and preventative maintenance and procedural surveillance histories related to the boric acid filter to verify that the design bases were appropriately implemented and maintained in relation to the monitoring, operation, and maintenance of the filter system. The team reviewed records of corrective and preventative maintenance and applicable corrective actions to verify that potential degradation was being identified, monitored, and corrected, and that filter change outs were conducted at criteria consistent with the filter system design. The team also conducted a field walkdown with the system engineer of plant features associated with the boric acid filter and its design purpose to verify, by visual observation of reasonably accessible locations, that the installed configuration and material condition of the features are consistent with the design basis and plant drawings.

b. Findings

No findings of significance were identified.

.2.15 Reactor Trip Breakers (RTB)

a. Inspection Scope

The team reviewed the RTB DBDs, UFSAR, and applicable plant drawings to identify the design basis function of the RTB. The team verified by review of schematic drawings, that the operation of the breakers was consistent with the design basis. Also, the team reviewed calculations to verify circuit breaker ratings and protective devices trip settings for consistency with the operational requirements. The team reviewed machinery history for failures. In addition, the team reviewed and evaluated the significance of selected corrective action documents related to RTB to verify problems were adequately resolved.

b. Findings

No findings of significance were identified.

.2.16 Reactor Vessel Level Instrumentation System (RVLIS)

a. Inspection Scope

The team reviewed calculations to verify that the design bases and design assumptions have been appropriately translated into design calculations and procedures. The team reviewed the system health and status report to determine the system equipment performance, reliability, maintenance, and material indicators/issues. The team reviewed the instrumentation setpoints, supporting uncertainty calculations/bases, surveillance/calibration procedures, channel calibration procedures, and calibration intervals for the electronic transmitters to ensure technical adequacy. The team reviewed the Westinghouse vendor instruction manual for RVLIS to verify compliance with vendor recommendations. The team reviewed NRC Information Notice (IN) 97-25, Dynamic Range Uncertainties in the Reactor Vessel Level Instrumentation to verify the licensees the response/applicability determination. The team reviewed the operations training procedures for RVLIS and maintenance qualification requirements for maintaining RVLIS to verified that instrumentation, alarms and instructions were available to operators for making necessary decisions.

b. Findings

No findings of significance were identified.

.2.17 Switchyard Batteries Used for Breaker Recovery

a. Inspection Scope

The team reviewed calculations to verify the switchyard primary and backup 125V DC batteries and chargers for sizing, loading, short-circuit, and voltage drop. The team reviewed the primary and backup switchyard battery maintenance procedures and last maintenance report to determine the licensees identification of problems. The team reviewed the ground detection instrumentation and maintenance procedures to determine the licensees identification of problems. The team performed a visual non-intrusive inspection of the switchyard primary and backup batteries and chargers to assess the installation configuration, material condition, and potential vulnerability to hazards. The team reviewed the transmission issued action guidelines for establishing emergency DC power to 230kV generator breakers 52-8 or 52-9 following a loss of station batteries and performed a walkdown of the steps outlined in the breaker recovery guideline to ensure the actions can be achieved.

b. Findings

No findings of significance were identified.

.2.18 Station Batteries

a. Inspection Scope

The team reviewed electrical calculations for the safety related station batteries A and B including battery duty cycle and voltage drop calculations, short circuit fault current calculation, breaker interrupting ratings and electrical coordination for accuracy. The team reviewed electrical calculations for the battery float and equalizing voltages, overall battery capacity for accuracy. The team reviewed five-year performance discharge test and quarterly battery surveillance tests to verify acceptance criteria. In addition, the voltage drop calculations for safety-related DC loads and DC control power to 480V switchgear was reviewed to verify if adequate voltage was available at these loads during the first hour of the station blackout event. The team reviewed calculations to verify minimum and maximum battery room temperatures were consistent with design basis requirements. Also, the team reviewed the manufacturing date codes on the battery cells to establish the age of the battery and assess whether any cells have been replaced. The team performed a walkdown of the battery station to assess observable material condition.

b. Findings

No findings of significance were identified.

.2.19 Battery Chargers

a. Inspection Scope

The team reviewed electrical design documents for 125 Vdc battery charger, including sizing calculation, its contribution to short circuit fault current, and breaker sizing to verify accuracy. In addition, the test procedures were reviewed to determine if maintenance and testing activities for the battery chargers were in accordance with UFSAR requirements and vendor recommendations. Also, the team performed a walkdown of the battery chargers to verify the as-built configuration and assess their observable material condition.

b. Findings

No findings of significance were identified.

.2.20 Emergency Diesel Generator - Electrical Subsystems

a. Inspection Scope

The team reviewed energy sources, including those used for control functions, that would be available to verify the adequacy during required design bases conditions. The team reviewed instrumentation and alarms to verify availability to operators for making necessary decisions. The team reviewed calculations to verify that design bases and design assumptions have been appropriately translated into the design calculations and procedures. The team reviewed calculations to verify that the EDG was adequately protected with properly set protective devices during test mode and emergency operation including short-circuit capability of the output breaker under worst fault conditions. The team reviewed analyses/testing to assess EDG operation under required operating conditions. The team reviewed calculations and assessments to verify that: 1) steady-state and transient loading are within design capabilities, 2)adequate voltage would be present to start and operate connected loads, and 3)operation at maximum allowed frequency would be within the design capabilities. The team reviewed the DC control circuit loop analysis associated with the EDG breaker trip/close circuits to ensure adequate control voltage would be available. The team reviewed the basis for the EDG load sequence time delay setpoints, calibration intervals, and results of last calibration for accuracy. The team reviewed the EDG feeder breaker maintenance and controls to verify that the components will function when required. The team reviewed the interfaces and interlocks associated with the 480V switchgear E1 and E2, including voltage protection schemes (degraded voltage and loss of voltage relaying) that initiates connection to the EDG to verify adequacy. The team reviewed the setpoint calculations, calibration procedures, and latest surveillance results, for the voltage detection relays, including applicable time delays and the EDG breaker permissive voltage relay for accuracy and identification of problems. The team reviewed recently issued NRC IN 2007-27, Recurring Events involving Emergency Diesel Generator Operability, for applicability. The team performed a visual non-intrusive inspection of the emergency diesel generators to assess the installation configuration, material condition, and potential vulnerability to hazards.

b. Findings

The team identified that the licensee purchasing group that orders degraded grid undervoltage relays did not reference the correct specification for the relays. The relays installed may not be the relays that were specified for the installation. The relays (time delay element) that were specified to be purchased and installed were + 3% accuracy relays. This specification was not translated into the purchase order and + 10%

accuracy relays may have been ordered and installed. The relays are required to pass TS surveillance acceptance criteria of + 7%. The relays have been in service for the past 16 years and have passed the surveillance requirements. The concern is that if the relays installed are + 10% accuracy relays, after a seismic event the maximum drift could occur and the relays would subsequently be outside the required acceptance criteria for the designed application. The item is unresolved pending the licensees analysis to assess the acceptability of the installed relays and NRC review of this analysis. This issue was entered into the licensees corrective action program as NCR 241618. This issue is identified as an Unresolved Item (URI), URI 05000261/2007006-02, Incorrect Degraded Grid Undervoltage Relays Installed.

.3 Review of Low Margin Operator Actions

a. Inspection Scope

The team performed a margin assessment and detailed review of six risk significant and time critical operator actions. Where possible, margins were determined by the review of the assumed design basis and UFSAR response times and performance times documented by job performance measures (JPMs). For the selected components and operator actions, the team performed an assessment of the EPPs, Abnormal Operating Procedures (AOPs), Alarm Response Procedures (ARPs), and other operations procedures to determine the adequacy of the procedures and availability of equipment required to complete the actions. Operator actions were observed on the plant simulator and during plant walkdowns.

The following operator actions were observed on the licensees operator training simulator:

• Restoring RHR flow during reduced inventory operations, AOP-020, Loss of Residual Heat Removal (Shutdown Cooling).

• Initiating Emergency Boration of the RCS, FRP-S.1, Response to Nuclear Power Generation/ATWS.

• Loss of ultimate heat sink, EPP-28 Loss of Ultimate Heat Sink.

• Actions on loss of normal AFW suction source, EPP-4, Reactor Trip Response.

Additionally, the inspectors walked down, table-topped and investigated the following operational scenarios:

• Venting an RHR pump, AOP-020, Attachments 1 and 2.

• Manual actions to establish emergency cooling to charging pumps, AOP-014, Component Cooling Water System Malfunction, Attachment 1.

• Aligning alternate sources for AFW, OP-402, Auxiliary Feedwater System

• Local operation to restart battery chargers following trip, EPP-1, Loss of All AC Power.

• Establishing deep well cooling per EPP-28, Attachment 6.

• Manual actions to energize plant equipment using the Dedicated Shutdown Diesel Generator (DSDG), EPP-25, Energizing Supplemental Plant Equipment Using the DSDG.

b. Findings

No findings of significance were identified.

.4 Review of Industry Operating Experience

a. Inspection Scope

The team reviewed selected operating experience issues that had occurred at domestic and foreign nuclear facilities for applicability at the H. B. Robinson Nuclear Plant. The team performed an independent applicability review for issues that were identified as applicable to the H. B. Robinson Nuclear Power Plant and were selected for a detailed review. The issues that received a detailed review by the team included:

• NRC IN 2006-17, Recent Operating Experience of Service Water Systems Failures due to External Conditions.

• NRC IN 2005-11, Internal Flooding/Spray-down of Safety-Related Equipment due to Unsealed Equipment Hatch Floor Plugs and/or Blocked Floor Drains.

• NRC GL 89-13, Service Water System Problems Affecting Safety Related Equipment.

• NRC IN 1997-69, Reactor trip breakers and surveillance testing of auxiliary contacts.

• NRC IN 95-03, Loss of reactor coolant inventory and potential loss of emergency mitigation functions while in a shutdown condition.

• NRC IN 96-12, Control rod insertion problems.

b. Findings

No findings of significance were identified.

.5 Review of Permanent Plant Modifications

a. Inspection Scope

The team reviewed two modifications related to the selected risk significant components in detail to verify that the design bases, licensing bases, and performance capability of the components have not been degraded through modifications. The adequacy of design and post modification testing of these modifications was reviewed by performing activities identified in IP 71111.17, Permanent Plant Modifications, Section 02.02.a.

Additionally, the team reviewed the modifications in accordance with IP 71111.02, Evaluations of Changes, Tests, or Experiments, to verify the licensee had appropriately evaluated them for 10 CFR 50.59 applicability. The following modifications were reviewed:

• EC 59037, Install Deep Well Pump D.

• EC 64363, "A" CCW Pump Replacement Motor, cleanup revision following implementation, 01/19/07.

b. Findings

No findings of significance were identified.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On August 16, 2007, the team presented the inspection results to Mr. Walt and other members of the licensee staff. The team returned all proprietary information examined to the licensee. No proprietary information is documented in the report.

4OA7 Licensee-Identified Violations

The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meet the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a Non-Cited Violation.

TS 5.4.1 requires that written procedures shall be established, implemented, and maintained covering activities that include a loss of shutdown cooling. Contrary to this, on July 17, 2007, the licensee determined that procedure AOP-020, was inadequate in that it did not ensure operators would complete RHR pump venting prior to restarting an RHR pump when restoring shutdown cooling following an RCS leak while in reduced inventory if Core Exit Thermocouples (CET) were greater than 200 EF.

This finding is more than minor because it affects the procedure quality attribute of the reactor safety/mitigating systems cornerstone. It impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. In accordance with NRC Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significant Determination Process, and determined the finding was of very low safety significance (Green). This outcome was primarily due to the strength of the operator training that is in place that would deter the operators from starting a pump while venting was in progress. AOP-020 directed the venting of the RHR pumps; however, the procedure did not ensure that venting was complete prior to moving on to the step that would start the RHR pump. This procedure inadequacy would be encountered during a loss of shutdown cooling due to a leak while at reduced inventory if CET were greater than 200 EF. This finding is not greater than green because related operator training and licensees standard error prevention techniques would have likely prevented actual pump starting under such conditions. The licensee has revised the procedure to address the inadequacy.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

T. Walt, Site Vice President

C. Baucom, Licensing Supervisor

T. Tovar, Operations Manager

E. Caba, Design Superintendent

G. Sanders, Licensing

R. Supler, Supervisor Electrical and I&C Design

P. Fagan, Mechanical/Civil Design Supervisor

B. Stover, Ops/Work Control SRO

NRC

R. Hagar, Senior Resident Inspector

L. Cain, RII, Engineering Branch 1, Acting Chief

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000261/2007006-02 URI Incorrect Degraded Grid Undervoltage Relays Installed. (Section 1R21.2.20)

Closed

05000261/2006005-01 URI Equipment Performance for Functional Recovery After Certain UHS Scenarios.

(Section 1R21.2.9)

Opened and Closed

05000261/2007006-01 FIN Failure to install Thermal Overload (TOL)

protection on the D deep well pump.

(Section 1R21.2.9)