IR 05000161/2005003

IR 05000161-05-003 on 3/1/2005 - 6/30/2005 for H.B. Robinson, Unit 2; Permanent Plant Modifications, Post Maintenance Testing

ML052100409
Person / Time
Site:	Robinson, 05000161
Issue date:	07/29/2005
From:	Fredrickson P NRC/RGN-II/DRP/RPB4
To:	Moyer J Carolina Power & Light Co
References
IR-05-003
Download: ML052100409 (35)

Text

uly 29, 2005

SUBJECT:

H.B. ROBINSON STEAM ELECTRIC PLANT - NRC INTEGRATED INSPECTION REPORT 05000261/2005003

Dear Mr. Moyer:

On June 30, the US Nuclear Regulatory Commission (NRC) completed an inspection at your H.B. Robinson reactor facility. The enclosed integrated inspection report documents the inspection findings, which were discussed on July 6, with you and other members of your staff, and on July 20, with Mr. Bill Noll 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.

This report documents one finding concerning appropriate acceptance criteria not being included in two procedures for restoration of cooling to the reactor coolant pump seals following a loss of all seal cooling. The finding has potential safety significance greater than very low significance; however, it does not represent an immediate safety concern. In addition, the report documents one issue of very low safety significance (Green). This issue was determined to involve a violation of NRC requirements. However, because of its very low safety significance and because it has been entered into your corrective action program, the NRC is treating this issue as a non-cited violation (NCV), in accordance with Section VI.A.1 of the NRCs Enforcement Policy. If you contest this NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the H.B. Robinson facility.

CP&L 2 In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's 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/

Paul E. Fredrickson, Chief Reactor Projects Branch 4 Division of Reactor Projects Docket No.: 50-261 License No.: DPR-23

Enclosure:

Inspection Report 05000261/2005003 w/Attachment: Supplemental Information

REGION II==

Docket No: 50-261 License No: DPR-23 Report No: 005000261/2005003 Facility: H. B. Robinson Steam Electric Plant, Unit 2 Location: 3581 West Entrance Road Hartsville, SC 29550 Dates: April 1, 2005 - June 30, 2005 Inspectors: R. Hagar, Senior Resident Inspector D. Jones, Resident Inspector L. Miller, Senior Emergency Preparedness Inspector (Sections 1EP1, 1EP4, 4OA1)

J. Kreh, Emergency Preparedness Inspector (Sections 1EP1 and 4OA1)

N. Sanfilippo, Emergency Preparedness Specialist (Section 1EP1)

Approved by: P. Fredrickson, Chief Reactor Projects Branch 4 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

IR 05000261/2005003, Carolina Power and Light Company; on 3/1/2005 - 6/30/2005 ; H.B.

Robinson Steam Electric Plant, Unit 2; Permanent Plant Modifications, Post Maintenance Testing The report covered a three-month period of inspection by resident inspectors and an announced inspection by regional emergency preparedness inspectors. One Green non-cited violation, and one unresolved item with potential safety significance greater than Green, were identified. 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 NRC's 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 inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, for failure to promptly identify and correct a condition adverse to quality, in that a surveillance procedure that directed unacceptable preconditioning of the residual heat removal (RHR) pumps was not identified and corrected from 1997 to 2005.

The finding is greater than minor because it is associated with the procedure quality attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective of ensuring the availability and reliability of systems that respond to events to prevent undesirable consequences. Unacceptable preconditioning of the RHR pump could mask a condition that renders the pump inoperable. The finding is of very low safety significance because it is not a design or qualification deficiency, does not represent an actual loss of safety function for a system or train, and is not risk significant due to a seismic, fire, flooding, or severe weather initiating event. (Section 1R19)

Cornerstone: Barrier Integrity

TBD. The inspectors identified an unresolved item involving two examples of a violation of 10 CFR 50, Appendix B, Criterion V, Procedures, for failure to include appropriate acceptance criteria in two procedures for restoration of cooling to the reactor coolant pump seals following a loss of all seal cooling. The finding has potential safety significance greater than very low significance and requires the completion of a significance determination process Phase 3 review.

Both examples of the finding are greater than minor because they are associated with the procedure quality attribute of the Barrier Integrity Cornerstone and affect the cornerstone objective of providing reasonable assurance that the reactor coolant system protects the public from radionuclide releases caused by accidents or events. The examples are also determined to potentially have greater than very low safety significance, because, since the finding affects a Barrier Integrity Cornerstone objective, the NRC Significance Determination Process Phase 1 worksheet requires a Phase 3 risk evaluation be completed. This evaluation could result in the finding having significance greater than very low safety significance.(Section 1R17)

Licensee-Identified Violations

None.

REPORT DETAILS

Summary of Plant Status The unit began the inspection period at full rated thermal power. On May 13, power was reduced to approximately 50 percent power to enable repair of a steam leak inside containment. Approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> later, on May 14, the unit was returned to full power, and operated at full power for the remainder of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R04 Equipment Alignment

a. Inspection Scope

Partial System Walkdowns The inspectors performed the following three partial system walkdowns, while the indicated structures, systems, and/or components (SSCs) were out-of-service for maintenance and testing, System Walked Down SSC Out of Service Date Inspected B and C Charging Pump A Charging Pump May 2 B Emergency Diesel A Emergency Diesel May 10 Generator Generator Safety Injection Train A A Safety Injection Pump June 9 To evaluate the operability of the selected trains or systems under these conditions, the inspectors compared observed positions of valves, switches, and electrical power breakers to the procedures and drawings listed in the Attachment.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

a. Inspection Scope

For the six areas identified below, the inspectors reviewed the control of transient combustible material and ignition sources, fire detection and suppression capabilities, fire barriers, and any related compensatory measures to verify that those items were consistent with Updated Final Safety Analysis Report (UFSAR) Section 9.5.1, Fire Protection System, and UFSAR Appendix 9.5.A, Fire Hazards

Analysis.

The inspectors walked down accessible portions of each area and reviewed results from related surveillance tests to verify that conditions in these areas were consistent with descriptions of the areas in the UFSAR. Documents reviewed are listed in the

.

The following areas were inspected:

Fire Zone Description Unit 2 Cable Spreading Room Auxiliary Building Hallway (Ground Floor)

Pipe Alley Control Room Service Water Pump Area Diesel Generator B Room

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures

a. Inspection Scope

External Flooding The inspectors walked down the 226-foot elevation of the auxiliary building which contains risk-significant SSCs which are susceptible to flooding from external sources to verify that the area configuration, features, and equipment functions were consistent with the descriptions and assumptions used in FSAR Section 3.4, Water Level (Flood)

Design, and in the supporting basis documents listed in the Attachment. The inspectors reviewed the operator actions described in Procedure OMM-021, Operation During Adverse Weather, to verify that the desired results could be achieved.

Internal Flooding Because the emergency diesel generator (EDG) rooms contain risk-significant SSCs which are susceptible to flooding from postulated pipe breaks, the inspectors walked down the EDG rooms to verify that the area configuration, features, and equipment functions were consistent with the descriptions and assumptions used in Calculation RNP-F/PSA-0009, Assessment of Internally Initiated Flooding Events, and in the supporting basis documents listed in the Attachment. The inspectors reviewed the operator actions credited in the analysis to verify that the desired results could be achieved using the plant procedures listed in the Attachment.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification

a. Inspection Scope

The inspectors observed licensed-operator performance during requalification simulator training for crew 1 to verify that operator performance was consistent with expected operator performance, as described in the Continuing Training Simulator Option Form dated 5-9-05. This training tested the operators ability to respond to a security event which disabled important safety-related equipment, and to respond to a loss of the plants ultimate heat sink. The inspectors focused on clarity and formality of communication, the use of procedures, alarm response, control board manipulations, group dynamics, and supervisory oversight. Documents reviewed are listed in the

.

The inspectors observed the post-exercise critique to verify that the licensee identified deficiencies and discrepancies that occurred during the simulator training.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed the two degraded SSC/function performance problems or conditions listed below to verify the appropriate handling of these performance problems or conditions in accordance with 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, and 10 CFR 50.65, Maintenance Rule. Documents reviewed are listed in the

.

The problems/conditions and their corresponding action requests (ARs) were:

Performance Problem/Condition AR

[Level indicator]-496, C steam generator narrow-range level, was 121479 indicating 12 percent while the recorder and the redundant channels were indicating the program level of approximately 53 percent

[Containment cooling fan] HVH-1 inboard bearing showed an increasing 113595 trend in vibration levels During the reviews, the inspectors focused on the following:

• Appropriate work practices,

• Identifying and addressing common cause failures,

• Scoping in accordance with 10 CFR 50,65(b),

• Characterizing reliability issues (performance),

• Charging unavailability (performance),

• Trending key parameters (condition monitoring),

• 10 CFR 50,65(a)(1) or (a)(2) classification and reclassification, and

• Appropriateness of performance criteria for SSCs/functions classified (a)(2) and/or appropriateness and adequacy of goals and corrective actions for SSCs/functions classified (a)(1).

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Evaluation

a. Inspection Scope

For the four time periods listed below, the inspectors reviewed risk assessments and related activities to verify that the licensee performed adequate risk assessments and implemented appropriate risk-management actions when required by 10 CFR 50.65(a)(4). For emergent work, the inspectors also verified that any increase in risk was promptly assessed, and that appropriate risk-management actions were promptly implemented. Documents reviewed are listed in the Attachment. Those periods included the following:

• April 22 - April 28, including scheduled maintenance on the B motor-driven auxiliary feedwater pump

• May 16 - May 20, including scheduled maintenance on the steam-driven auxiliary feedwater pump and the startup transformer

• May 30 - June 3, including scheduled maintenance on the B residual heat removal train and component cooling water pumps

• June 27 - June 30, including emergent maintenance on the C charging pump and D instrument air compressor

b. Findings

No findings of significance were identified.

1R14 Personnel Performance During Nonroutine Plant Evolutions

a. Inspection Scope

During the two non-routine plant evolutions described below, the inspectors observed plant instruments and operator performance to verify that the operators performed in accordance with the associated procedures and training.

• The downpower from 100 percent to approximately 50 percent power on May 13

• The subsequent return to full power on May 14 Documents reviewed are listed in the Attachment.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed two operability determinations as described below.

• The inspectors reviewed the operability determination associated with AR 157455.

This AR addressed the effects of a small steam leak on the operability of a containment penetration. The inspectors assessed the accuracy of the evaluation, the need for any associated compensatory measures, and compliance with the technical specifications (TS). The inspectors also verified that the operability determination was completed as specified by Procedure PLP-102, Operability Determinations.

• The inspectors reviewed Engineering Change 52332, [Steam Generator A Steam Flow Transmitter]-475 Replacement, a deferred modification, to verify that this modification did not involve a degraded component for which an operability evaluation was warranted.

Documents reviewed are listed in the Attachment.

b. Findings

No findings of significance were identified.

1R16 Operator Work-Arounds

a. Inspection Scope

The inspectors reviewed Work-Around 05-03, Ground on 480 Volt Bus 5 Will Not Alarm, to verify that this work-around did not affect either the functional capability of the related system in responding to an initiating event, or the operators ability to implement abnormal or emergency operating procedures.

The inspectors reviewed the cumulative effects of the operator work-arounds that were in place on March 31 to verify that those effects could not increase an initiating event frequency, affect multiple mitigating systems, or affect the ability of operators to respond in a correct and timely manner to plant transients and accidents. The work-arounds included in this review included the following:

• 04-06, [Flow element]-660 reads zero flow

• 04-11, Additional time is needed to start or stop Deepwell Pump B

• 04-12, Caustic dilution flow is reduced to the makeup water treatment primary mixed bed demineralizers

• 04-13, [Pressure control valve]-1380, main steam from left main steam stop valves to high-pressure turbine, leaks by its seat

• 04-14, Spurious alarm on reactor coolant pump thermal barrier low flow

• 04-15, Flow switches for air handlers HVE-6A & HVE-6B are defeated

• 04-18, Postulated fire in zone 19 or 20 may cause spurious closing of [level control valve]-115C and loss of [level control valve]-115B function Documents reviewed are listed in the Attachment.

b. Findings

No findings of significance were identified.

1R17 Permanent Plant Modifications

.1 Restoration of RCP Seal Cooling

a. Inspection Scope

After noting that a previously-inspected modification had included a design requirement to restore cooling to the RCP seals within 13 minutes following a loss of all seal cooling caused by a particular event, the inspectors reviewed other licensee procedures which restore seal cooling following a loss of all seal cooling caused by other events, to verify that those procedures were consistent with that design requirement. (The previously-inspected modification was Engineering Change 59037, described in NRC Inspection Report 05000261/2005002.)

b. Findings

Introduction The inspectors identified an unresolved item involving two examples of a violation of 10 CFR 50, Appendix B, Criterion V, Procedures, for failure to include appropriate acceptance criteria in two procedures for restoration of cooling to the reactor coolant pump (RCP) seals following a loss of all seal cooling. The finding has potential safety significance greater than very low significance and requires the completion of a significance determination process Phase 3 review.

Description Several industry documents indicate that restoration of RCP seal cooling is time-sensitive, in that restoration of seal coolant after RCP temperature becomes elevated could cause additional pump damage, such as seal degradation and increased seal leakage. Additionally, the Westinghouse Owners Group recommended that RCP seal cooling not be restored following a prolonged loss of seal cooling in which the RCP seal temperature becomes elevated. Thus, restoration of seal cooling would be the proper action to take, only if seal cooling is restored before hot RCS fluid reaches the RCS seals and causes the seal temperature to increase high enough to create seal degradation and subsequent leakage.

Procedure EPP-22, Revision 20, Energizing Plant Equipment Using Dedicated Shutdown Diesel Generator, provides instructions for restoring RCP seal cooling and other evolutions following a loss of all alternating-current power, but does not include any requirement or precaution regarding the time at which RCP seal cooling is restored.

Therefore, EPP-22 does not include an appropriate criterion to ensure that restoration of seal cooling is accomplished before hot RCS fluid reaches the RCP seals.

Procedure DSP-002, Revision 30, Hot Shutdown Using the Dedicated/Alternate Shutdown System, provides instructions for establishing and maintaining hot-shutdown conditions following a fire that precludes the use of emergency operating procedures from the main control room. Among the evolutions directed by DSP-002 is the restoration of RCP seal cooling. However, DSP-002 also does not include any requirement or precaution regarding the time at which RCP seal cooling is restored.

Therefore, DSP-002 does not include an appropriate criterion to ensure that restoration of seal cooling is accomplished before hot RCS fluid reaches the RCP seals.

Analysis Both examples of the finding are greater than minor because they are associated with the procedure quality attribute of the Barrier Integrity Cornerstone and affect the cornerstone objective of providing reasonable assurance that the reactor coolant system protects the public from radionuclide releases caused by accidents or events. The examples are also determined to potentially have greater than very low safety significance, because, since the finding affects a Barrier Integrity Cornerstone objective, the NRC Significance Determination Process Phase 1 worksheet requires a Phase 3 risk evaluation be completed. This evaluation could result in the finding having significance greater than very low safety significance.

Enforcement 10 CFR 50, Appendix B, Criterion V, Procedures, requires, in part, that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances. It further requires that these procedures include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Contrary to the above, Procedures EPP-22, Rev. 20 and DSP-002, Rev. 30, do not include appropriate acceptance criteria for restoration of cooling to the RCP seals following a loss of all seal cooling.

Pending determination of safety significance, this finding, with two examples, is identified as URI 05000261/2005003-01, Failure of Two Procedures to Have Appropriate Acceptance Criteria for Restoration of RCP Seal Cooling. This finding is in the corrective action program (CAP) as AR 160309.

.2 Modification Review

a. Inspection Scope

The inspectors reviewed the following two modifications:

The inspectors reviewed these modifications to verify that:

• the modifications did not degrade the design bases, licensing bases, and performance capabilities of risk significant SSCs,

• implementing these modifications did not place the plant in an unsafe condition, and

• the design, implementation, and testing of these modifications satisfied the requirements of 10 CFR 50, Appendix B.

Documents reviewed are listed in the Attachment.

b. Findings

No findings of significance were identified.

1R19 Post Maintenance Testing

a. Inspection Scope

For the five post-maintenance tests listed below, the inspectors witnessed the test and/or reviewed the test data to verify that test results adequately demonstrated restoration of the affected safety functions described in the UFSAR and TS.

Documents reviewed are listed in the Attachment.

The following tests were witnessed/reviewed:

Related Test Procedure Title Maintenance Activity Date Inspected OST-101-7 Comprehensive Flow Replace valves and April 13 Test for Charging speed controller on the Pump B B charging pump OST-202 Steam Driven Inspect the impeller on May 17 Auxiliary Feedwater the steam driven System Component auxiliary feedwater Test pump OST-252-2 [Residual Heat Routine maintenance June 1 Removal] System on residual heat Valve Test - Train B removal train B OST-151-1 Safety Injection Replace the motor for June 8 System Components safety injection pump Test - Pump A A OST-201-1 [Motor-Driven Calibrate pressure June 13 Auxiliary Feedwater] gauges and repair a System Component packing leak Test - Train A

b. Findings

Introduction The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, for failure to promptly identify and correct a condition adverse to quality, in that a surveillance procedure that directed unacceptable preconditioning of the residual heat removal (RHR) pumps was not identified and corrected from 1997 to 2005.

Description In 1997, the licensee entered NRC Information Notice (IN) 97-16 into the Operating Experience Assessment program as OEA 6293. That IN discussed preconditioning issues, and specifically identified venting an RHR pump without adequate controls immediately before testing the pump, as an example of unacceptable preconditioning. NUREG 1482, Guidelines for Inservice Testing at Nuclear Plants; and NRC Inspection Manual, Part 9900: Technical Guidance, Maintenance - Preconditioning of Structures, Systems, and Components Before Determining Operability also state that routine uncontrolled pump venting directly preceding surveillance testing without proper controls is unacceptable preconditioning, because this practice could mask a condition that could have rendered the pump inoperable.

In April, 1997 the licensee initiated AR 97-00847 to address instances of preconditioning valves prior to performing surveillance testing. Through the associated investigation, the licensee developed corrective actions to evaluate all surveillance tests for preconditioning concerns. In September, 1997, the licensee closed OEA 6293 stating that although preconditioning was applicable to Robinson, no evaluation was required because AR 97-00847 addressed the issue. The licensee subsequently closed AR 97-00847. However, the corrective actions for AR 97-00847 failed to identify and correct surveillance test Procedure OST- 251-2, Residual Heat Removal Pump B and Components Test, which directed the operators to vent the pump immediately prior to testing but did not impose any controls on the venting. As a result, the licensee continued the practice of unacceptably preconditioning the RHR pumps from 1997 through 2005.

Analysis The finding is greater than minor because it is associated with the procedure quality attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective of ensuring the availability and reliability of systems that respond to events to prevent undesirable consequences. Unacceptable preconditioning of the RHR pump could mask a condition that renders the pump inoperable. The inspectors determined that the finding is of very low safety significance because it is not a design or qualification deficiency, does not represent an actual loss of safety function for a system or train, and is not risk significant due to a seismic, fire, flooding, or severe weather initiating event.

Enforcement 10CFR50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that conditions adverse to quality shall be promptly identified and corrected.

Contrary to the above, corrective actions in response to IN 97-16 and AR 97-00847 failed to identify and correct Procedure OST- 251-2, thus allowing unacceptable preconditioning of the RHR pumps to continue from 1997 to 2005. Because this failure to promptly identify and correct this inadequate procedure is of very low safety significance, and was entered into the CAP as AR 160962, it is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000261/2005-02

1R22 Surveillance Testing

a. Inspection Scope

For the five surveillance tests listed below, the inspectors witnessed testing and/or reviewed the test data to verify that the systems, structures, and components involved in these tests satisfied the requirements described in the TS, the UFSAR, and applicable licensee procedures, and that the tests demonstrated that the SSCs were capable of performing their intended safety functions. Documents reviewed are listed in the

.

Test Procedure Title Date Inspected OST-256 [Residual Heat Removal] Pump Pit April 19 Instrumentations Check Valve Back Flow Testing (Annual)

OST-401-2 [Emergency Diesel Generator] B Slow April 25 Speed Start OST-701-8* V12-10 and V12-11 Inservice Valve May 11 Test OST-901** [Heating Ventilation Recirculation] May 15 Condensate Measuring System (Weekly)

OST-352-2 Containment Spray Component Test - June 23 Train B

• This procedure included inservice testing requirements.

• • This procedure was a Reactor Coolant System leakage detection surveillance.

b. Findings

No findings of significance were identified.

1R23 Temporary Plant Modifications

a. Inspection Scope

The inspectors reviewed the temporary modification described in Engineering Change 61452, Inhibit [Radiation Monitor]-19C [Steam Generator] Blowdown Isolation Function to verify that the modification did not affect the safety functions of important safety systems, and to verify that the modification satisfied the requirements of Procedure EGR-NGGC-005, Engineering Change, and 10 CFR 50, Appendix B, Criterion III, Design Control.

Documents reviewed are listed in the Attachment.

b. Findings

No findings of significance were identified.

Cornerstone: Emergency Preparedness

1EP1 Exercise Evaluation

a. Inspection Scope

Prior to the onsite inspection, an in-office review was conducted of the exercise objectives and scenario submitted to the NRC to verify that the exercise would test major elements of the Emergency Plan as required by 10 CFR 50.47(b)(14).

The onsite inspection consisted of reviewing and assessing the following items:

• The adequacy of the licensees performance in the biennial exercise was reviewed and assessed regarding the implementation of the risk-significant planning standards in 10 CFR 50.47 (b) (4), (5), (9), and (10), which are emergency classification, offsite notification, radiological assessment, and protective action recommendations, respectively.

• The overall adequacy of the licensees emergency response facilities with regard to NUREG-0696, Functional Criteria for Emergency Response Facilities, and Emergency Plan commitments. The facilities assessed were the control room simulator, technical support center, and emergency operations facility.

• Other performance areas besides the risk significant planning standards, such as the emergency response organizations (ERO) recognition of abnormal plant conditions, command and control, intra- and inter-facility communications, prioritization of mitigation activities, utilization of repair and field monitoring teams, interface with offsite agencies, and the overall implementation of the emergency plan and its implementing procedures.

• Past performance issues from NRC inspection reports and Federal Emergency Management Agency exercise reports to determine effectiveness of corrective actions as demonstrated during this exercise to ensure compliance with 10 CFR 50.47(b)(14).

• The post-exercise critique to evaluate the licensees self-assessment of its ERO performance during the exercise and to ensure compliance with Section IV.F.2.g of Appendix E to 10 CFR Part 50.

Documents reviewed are listed in the Attachment.

b. Findings

No findings of significance were identified.

1EP4 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope

The inspectors review of revisions to the Emergency Plan, implementing procedures and emergency action level changes was performed to determine whether changes had decreased the effectiveness of the plan. The inspectors also evaluated the associated 10 CFR 50.54(q) reviews associated with non-administrative emergency plan, implementing procedure and emergency action level changes. The revisions 57 and 58 covered the period from December 6, 2004 to May 6, 2005. The inspection results were evaluated against criteria in 10 CFR 50.47(b)(4) and its related 10 CFR Part 50, Appendix E requirements; NUREG-0654, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Revision 1, and Regulatory Guide 1.101, Emergency Planning and Preparedness for Nuclear Power Reactors, Revision 4.

Documents reviewed are listed in the Attachment.

b. Findings

No findings of significance were identified.

1EP6 Drill Evaluation

a. Inspection Scope

On April 5, the inspectors observed an emergency preparedness drill to verify licensee self-assessment of classification, notification, and protective action recommendation development in accordance with 10 CFR 50, Appendix E. The inspectors also attended the post-drill critique to verify that the licensee properly identified failures in classification, notification and protective action recommendation development activities.

b. Findings

No findings of significance were identified.

4OA1 Performance Indicator (PI) Verification

a. Inspection Scope

To verify the accuracy of the PI data for the PIs evaluated below, PI definitions and guidance contained in NEI 99-02, Regulatory Assessment Indicator Guideline, Revision 2, were used to verify the basis in reporting for each data element.

Cornerstone: Emergency Preparedness

The inspectors reviewed the procedure for developing data for the Emergency Plan performance indicators, which are:

(1) Drill and Exercise Performance;

(2) Emergency Response Organization Drill Participation; and

(3) Alert and Notification System Reliability. The inspectors examined data reported to the NRC for the period October 2004 to March 2005. Procedural guidance for reporting PI information and records used by the licensee to identify potential PI occurrences were also reviewed.

• Drill and Exercise Performance The inspectors verified the accuracy of the PI for Drill and Exercise Performance through review of a sample of drill and event records.

• Emergency Response Organization Drill Participation The inspectors reviewed selected training records to verify the accuracy of the PI for ERO drill participation for personnel assigned to key positions in the ERO.

• Alert and Notification System Reliability The inspectors verified the accuracy of the PI for alert and notification system reliability through review of a sample of the records of periodic system tests.

Documents reviewed are listed in the Attachment.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

.1 Routine Review of ARs

To aid in the identification of repetitive equipment failures or specific human performance issues for followup, the inspectors performed frequent screenings of items entered into the CAP. The review was accomplished by reviewing daily ARs.

.2 Annual Sample Review

a. Inspection Scope

reviewed this report to verify:

• complete and accurate identification of the problem in a timely manner;

• evaluation and disposition of performance issues;

• evaluation and disposition of operability and reportability issues;

• consideration of extent of condition, generic implications, common cause, and previous occurrences;

• appropriate classification and prioritization of the problem;

• identification of root and contributing causes of the problem;

• identification of corrective actions which were appropriately focused to correct the problem; and

• completion of corrective actions in a timely manner.

The inspectors also reviewed this AR to verify licensee compliance with the requirements of the CAP as delineated in Procedure CAP-NGGC-0200, Corrective Action Program, and 10 CFR 50, Appendix B.

Documents reviewed are listed in the Attachment.

b. Observations and Findings

No findings of significance were identified.

.3 Semi-Annual Trend Review

a. Inspection Scope

The inspectors performed a review of the CAP and associated documents to identify trends that could indicate the existence of a more significant safety issue. The inspectors review focused on repetitive equipment issues, but also considered the results of daily inspector CAP item screening discussed in Section 4OA2.1, licensee trending efforts, and licensee human performance results. The inspectors review nominally considered the six month period of January 2005, through June 2005, although some examples expanded beyond those dates when the scope of the trend warranted. The review included issues documented outside the normal CAP in major equipment problem lists, repetitive and/or rework maintenance lists, departmental problem/challenges lists, system health reports, quality assurance audit/surveillance reports, self assessment reports, and Maintenance Rule assessments. The inspectors compared and contrasted their results with the results contained in the latest monthly and quarterly trend reports. Corrective actions associated with a sample of the issues identified in the trend reports were reviewed for adequacy. The specific documents reviewed are listed in the Attachment.

The inspectors also evaluated the trend reports against the requirements of the CAP as specified in 10 CFR 50, Appendix B, Criterion XVI, and in Procedures CAP-NGGC-0200, Corrective Action Program, and CAP-NGGC-0206, Corrective Action Program Trending and

Analysis.

b. Assessment and Observations No findings of significance were identified. The inspectors observed that through their trending methodology, the licensee had performed a detailed review. The licensee routinely reviewed cause codes, involved organizations, key words, and system links to identify potential trends in the CAP data. The inspectors compared the licensees process results with the results of the inspectors daily screening, and did not identify any discrepancies or potential trends in the CAP data that the licensee had failed to identify. However, the inspectors noted that Procedure CAP-NGGC-0206 states, in part, that the unit/section roll-up meetings are opportunities for line organizations to conduct cognitive trending, which the procedure defines as the process of maintaining a mental awareness of recent events and identifying trends via association of like items.

While reviewing the monthly trend reports, the inspectors noted that some personnel who attended unit/section roll-up meetings did not attend consistently. The inspectors considered that meeting participants would be able to perform cognitive trending in the unit/section roll-up meetings only if they attended those meetings, and that such trending would be most effective if meeting participants missed no meetings. Because some meeting participants had not consistently attended unit/section roll-up meetings, the inspectors concluded that the licensee had not been able to derive the full benefits of cognitive trending. To address this conclusion, the licensee initiated AR 161018.

4OA5 Other Activities

.1 Operational Readiness of Offsite Power (Temporary Instruction 2515/163)

The inspectors collected data from licensee maintenance records, corrective action documents and procedures, and through interviews of station engineering, maintenance, and operations staff, as required by TI 2515/163. Appropriate documentation of the inspection results was provided to headquarters staff for further analysis, as required by the TI. This completes the Region II inspection requirements for this TI for the Robinson site.

4OA6 Meetings, Including Exit

On July 6, the resident inspectors presented the inspection results to Mr. John Moyer and other members of his staff.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

R. Ivey, Operations Manager

C. Church, Engineering Manager

E. Caba, Engineering Superintendent

B. Clark, Nuclear Assurance Manager

J. Huegel, Maintenance Manager

W. Noll, Director of Site Operations

E. Kapopoulos, Outage Management Manager

D. Stoddard, Plant General Manager

W. Farmer, Engineering Superintendent

J. Lucas, Manager, Support Services - Nuclear

J. Moyer, Vice President, Robinson Nuclear Plant

A. Cheatham, Radiation Protection Superintendent

S. Wheeler, Supervisor, Regulatory Support

G. Ludlum, Training Manager

NRC personnel

P. Fredrickson, Chief, Reactor Projects Branch 4

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000261/2005003-01 URI Failure of Two Procedures to Have Appropriate Acceptance Criteria for Restoration of RCP Seal Cooling (Section 1R17)

Opened and Closed

05000261/2005003-02 NCV Failure to Identify and Correct a Surveillance Procedure That Unacceptably Preconditions the Residual Heat Removal Pumps (Section 1R19)

Closed

None

Discussed

None