IR 05000244/1999002

Insp Rept 50-244/99-02 on 990222-0404.No Violations Noted. Major Areas Inspected:Aspects of Licensee Operations,Maint, Engineering & Plant Support.Physical Security Program Met Safeguards Requirements

ML17265A654
Person / Time
Site:	Ginna
Issue date:	05/14/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML17265A653	List: IR 05000244/1999002 ML17265A652
References
50-244-99-02, 50-244-99-2, NUDOCS 9905210117
Download: ML17265A654 (46)

Text

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

License No.

DPR-18 Report No.

50-244/99-02 Docket No.

50-244 Licensee:

Facility Name:

Location:

Rochester Gas and Electric Corporation (RGSE)

R. E. Ginna Nuclear Power Plant 1503 Lake Road Ontario, New York 14519 Inspection Period:

Inspectors:

February 22, 1999 through April 4, 1999 P. D. Drysdale, Senior Resident Inspector C. C. Osterholtz, Resident Inspector E. H. Gray, Senior Reactor Inspector T. A. Moslak, Radiation Specialist P. R. Frechette, Physical Security Specialist Approved by:

G. Scott Barber, Acting Chief Projects Branch

Division of Reactor Projects 99052f0117 9905i4 PDR ADOCK 05000244

PDR

EXECUTIVE SUMMARY R. E. Ginna Nuclear Power Plant NRC Integrated Inspection Report 50-244/99-02

. This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support.

The report covers a 6-week period of resident inspection,

,and includes the results of announced inspections by regional specialists in the areas of inservice inspection, radiation protection, and physical security.

~Osra ions Control room operators responded well to anomalous plant conditions and performed well in controlling the plant during the shutdown and cooldown for a scheduled refueling outage.

The licensee made progress in improving system configuration controls, and the training conducted for licensed operators on past configuration control problems was a good initiative. However, several configuration control deficiencies occurred during the current refueling outage which indicate ongoing problems still existed in this area.

The new issues were entered into the licensee's corrective action program.

Main enance The licensee exercised systematic work practices and achieved good quality repairs during preventive maintenance inspections of plant equipment and circuit breakers.

Controlled procedures were in use at maintenance job sites, were up to date and were properly utilized by technicians involved in the outage work. The inspectors observed good personnel and plant safety practices during the maintenance work. A lower threshold for operability considerations during breaker maintenance had improved licensee identification and resolution of breaker problems.

Test activities involving safety injection accumulator check valves, emergency diesel generators, and the residual heat removal system were well controlled, and the systems were satisfactorily tested to assure operability prior to being returned to service.

fnservice inspection (ISI} activities were well planned and implemented by qualified personnel in accordance with approved procedures.

Inspector observation of nondestructive testing in progress showed that the ISI work was conducted with proper oversight by RG&E staff and the results were well documented.

The inspections observed were thorough and of sufficient extent to determine the integrity of the components inspected.

Problems were evaluated and effectively addressed in accordance with Code requirements.

The licensee made significant programmatic improvements in foreign material exclusion (FME) controls, and took actions to formally incorporate previous weaknesses into their corrective action process.

During the current refueling outage,.the licensee was able to identify causes for all of the FME incidents that had occurred to date, initiated corrective

Executive Summary (cont'd)

actions to recover the material, and implemented additional controls to prevent further occurrences.

n ineerin Plant modifications installed during the current refueling outage were good enhancements to the operation and reliability of plant equipment.

The installation packages reviewed contained detailed instructions and information for performing and documenting the modification work. The safety evaluations reviewed were adequate to demonstrate that the modifications did not represent any unreviewed safety concerns.

The licensee successfully resolved internal valve degradation that resulted from heavy throttling of a service water valve by replacing it with a smaller valve that was more resistant to erosion.

However, the system conditions that required heavy throttling of service water at the component cooling water heat exchangers were not yet resolved.

The licensee continued to evaluate the need to increase service water flow to reduce siltation and erosion, and to maintain optimal component cooling water system temperatures.

"As Low As Reasonably Achievable" (ALARA)program requirements were well developed, integrated in the work control process, and effectively implemented with respect to the in-service inspection of reactor components.

Dose levels received by individuals and work groups were closely monitored by the ALARAgroup.

Dose information was provided to

. management for timely resolution of emergent issues, resulting in cumulative doses below estimates.

The radiological controls program was effectively implemented by qualified and experienced staff properly implementing detailed procedures and radiation work permits, appropriatelg monitoring personnel, exposure, and adequately maintaining radiologically controlled areas.

Work performance standards were effectively monitored and reinforced by close and frequent management and quality assurance oversight.

Off-normal conditions were conservatively identified, appropriately evaluated, and resolved in a timely manner.

The licensee conducted security and safeguards activities in a manner that protected public health and safety in the areas of access authorization, alarm stations, communications, and protected area access control of personnel, packages and vehicles.

Security facilities and equipment were well maintained and reliable and were able to meet the licensee's commitments and NRC requirements.

Security force members had the requisite knowledge to effectively implement the duties and responsibilities of their position(s).

Management support was adequate to ensure effective implementation of the security program.

The licensee's audits were comprehensive in scope and depth, that the audit

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. findings were reported to the appropriate level of management, and that the program was being properly administered.

In addition, a review of the documentation applicable to the

Executive Summary (cont'd)

self-assessment program indicated that the program was being effectively implemented to identify and resolve potential weaknesse TABLE OF CONTENTS EXECUTIVE SUMMARY TABLE OF CONTENTS

. v I ~ Operations........ ~............

08 Conduct of Operations...................

~ 1 General Comments.......

Operational Status of Facilities and Equipment 02.1 Summary of Plant Status and Plant Shutdown for Refueling...

02.2 (Update) Inspector Follow-up Item (IFI) 50-244/97-10-01: Weak Configuration Control.......................

Miscellaneous Operations Issues...... ~...........

08.1 (Closed) Licensee Event Report (LER) 1999-002: "Surveillance Not Performed, Due to Personnel Error, Resulted in Violation of Technical Specifications.......

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M1 Conduct of Maintenance..............................

M1.1 General Maintenance Activities.......................

M1.2 General Surveillance Activities........................

M2 Maintenance and Material Condition of Facilities and Equipment.....

M2.1 Inservice inspection Work in Progress ~..................

M8 Miscellaneous Maintenance Issues M8.1 (Closed) IFI 50-244/97-11-01:

Foreign Material Exclusion (FME)

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M8,2 (Closed) IFI 50-244/98-07-01: Deficiencies in Work Planning and Spare Parts Requisitioning; Potential Inventory Control Problems

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E2 Engineering Support of Facilities and Equipment..............

E2.1 Installation of Plant Modifications E2.2 Service Water Valve V-4620 Inspection and Replacement E8 Miscellaneous Engineering Issues E8.1 (Closed) IFI 50-244/98-012-03:

Service Water Valve Testing E8.2 (Open) Licensee Event Report (LER) 1999-001: "Deficiencies in NSSS Vendor Steamline Break Mass and Energy Release Analysis Results in Plant Being Outside its Design Basis....

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R1 Radiological Protection and Chemistry (RPRC) Controls R1.1 Implementation of the Radiation Protection Program R2 Status of RPSC Facilities and Equipment R2.1 Radiological Work Practices, Access Controls, and Housekeeping R7.. Quality Assurance in RPSC Activities.............

R7.1 Audits, Surveillances, and Management Appraisals

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Table of Contents (cont'd)

RS S1 S2 SS S4 S5 S6 S7 Miscellaneous ROC Issues..........

R8.1 (Closed) IFI 50-244/97-11-04: Reactor Cavity/Refueling Canal Water Leakage Conduct of Security and Safeguards Activities S1.1 Security and Safeguards.. ~......,.........

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Status of Security Facilities and Equipment...................

S2.1 Security Facilities and Equipment........... ~.....

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Security and Safeguards Procedures and Documentation S3.1 Security Program Procedures................

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Security and Safeguards Staff Knowledge and Performance.......

S4.1 Staff Knowledge and Performance....................

Security and Safeguards Staff Training and Qualifications (TRQ)....

S5.1 Staff Training and Qualifications Security Organization and Administration....................

S6.1 Management Support and Effectiveness...........

Quality Assurance (QA) in Security and Safeguards Activities......

S7.1 Effectiveness of QA Activities.... ~..................

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XO INPO Plant Evaluation Report Review...............

X1 Exit Meeting Summary.........................

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~ 26 ATTACHMENTS Attachment 1-Partial List of Persons Contacted

- Inspection Procedures Used

- Items Opened, Closed, and Discussed

- List of Acronyms Used

Re ort Details I. 0 erations

Conduct of Operations'1.1

.General C mment Ins ec ion Pr c d re IP 71707 The inspectors periodically observed plant operations to verify that the facilitywas operated safely and in accordance with licensee procedures and regulatory requirements.

This review included tours of the accessible areas of the facility. The inspectors conducted ongoing verifications of engineered safety feature (ESF)

system operability, verifications of proper control room and shift staffing, verification that the plant was operated in conformance with the improved technical specifications (ITS) and appropriate action statements were implemented for out-of-service equipment, and verification that logs and records accurately identified equipment status or deficiencies.

Operational Status of Facilities and Equipment 02.1 Summa of Plant S atus and Plant S u down for Refuelin

-"The inspectors periodically observed operator performance and reviewed operational status of the plant throughout the inspection period.

b.

Observa ions and Findin s The plant was at approximately 76% power at the beginning of inspection period, in coastdown operations in preparation for a refueling outage.

On February 22, 1999, the licensee issued a four hour notification to the NRC after being informed by Westinghouse that the plant may have been outside of its design basis for a steam line break in containment (see section E8.2).

On February 25, 1999, calibrations were completed on the control room noble gas monitor (R-36), and the control room ventilation system was returned to a normal lineup.

However, on February 27, 1999, the licensee discovered a failed chlorine probe during maintenance activities and returned control room ventilation to the recirculation mode.

The ventilation system problems were resolved and control room ventilation was returned to a normal lineup on April 2, 1999.

Topical headings such as 01, M8, etc., are used in accordance with the NRC standardized reactor inspection report outline.

Individual reports are not expected to address all outline topic On February 27, 1999, the licensee issued a four hour notification to the NRC in accordance with 10 CFR 50.72 after an inadvertent containment ventilation isolation occurred while performing instrumentation and control (l&C) calibrations on

. the containment gas monitor (R-12).

The isolation signal.was later determined to be invalid due to a temporary jumper that did not maintain continuity. Although the jumper was adequately checked prior to its use, the licensee concluded that its design was not well suited for this application since space limitations prevented complete contact at the terminals.

The licensee later considered that the actuation signal was not valid and the notification was retracted on March 3, 1999.

On March 1, 1999, at 4:05 a.m., a planned shutdown and cooldown was commenced from approximately 6996 power to enter the 27th refueling outage.

Operators responded welt to several anomalies that occurred during the shutdown and cooldown.

The operators noted that the B-safety injection (B-Sl) channel could not be blocked for automatic SI until several pounds per square inch (psi) below the low primary system pressure setpoint.

The A-Sl channel was blocked normally.

The licensee generated an ACTION Report to address the issue (99-0239).

During the cooldown, operators also noted that condenser vacuum could not be maintained with one circulating water pump in service, and manually aligned the steam dumps to the atmospheric relief valves (ARVs) from the condenser when the low condenser pressure annunciator did not automatically do so (as designed).

The licensee generated two additional ACTION Reports (99-0240 and 99-0241) to address the vacuum and annunciator problems.

Improved Technical Specification (ITS) limiting

, condition for operation (LCO) 3.0.3 was briefly entered on March 'I, 1999, when the main steam check valves did not meet their acceptance criteria during testing (see section 08i3).

The plant entered MODE 5 on March 4, 1999, and MODE 6 was entered on March 5, 1999.

The reactor was defueled on March 11, 1999, and MODE 6 operations resumed April 3, 1999.

The plant was in MODE 6 performing refueling operations at the end of the inspection period.

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Conclu ions Control room operators responded well to anomalous plant conditions and performed well in controlling the plant during the shutdown and cooldown for a scheduled refueling outage.

02.2 at ns ec or Follow-u em IFI 50-244 97-10-01: W a nfi ra i n

~Cn rol ao fn ction Sc (7'I 707)

The inspectors reviewed the licensee's ongoing actions to improve plant configuration control b.

Observations and Findin s IFI 50-244/97-10-01 was opened during the 1997 refueling outage after multiple instances involving poor configuration controls occurredmhich led.to plant events.

In most cases, the deficiencies were licensee identified, and in all cases were entered into the licensee's corrective action program through the use of ACTION Reports.

Prior to the current refueling outage, licensee management conducted training on configuration controls with all the operating crews.

The training consisted of a review of all the ACTION Reports generated during the previous refueling outage that were related to configuration control problems, and a discussion of the human performance aspects associated with them.

The inspectors observed portions of the training for one crew and noted good presentations of the relevant issues, and good participation from crew members during the training.

During the current refueling outage, several configuration control problems occurred which indicated that the licensee continued to have difficulties in this area.

The first occurred on March 3, 1999, when the secondary plant retention tank was inadvertently overfilled while draining the condenser hotwell, and the turbine tube oil storage tank was inadvertently overflowed while transferring oil from the turbine lube oil reservoir.

The licensee generated two ACTION Reports (99-0261 and 99-0263) to document the issues, and operations management temporarily

. suspended auxiliary operator activities to discuss concerns with shift personnel.

The licensee also initiated an event investigation to determine the root causes and the human performance aspects of these events.

The investigation was completed on March 15, 1999, and concluded that inattention to detail resulted in incorrect assumptions being made by operations personnel, which in turn resulted in the events.

The inspectors reviewed the event investigation and determined that it was effectively performed in accordance'ith administrative procedures using appropriate personnel, and was completed in a timely manner.

The corrective actions for these events were still being evaluated at the end of the inspection period.

On March 4, 1999, while performing a drain down of the pressurizer to a level of 20 inches using a reactor coolant drain tank (RCDT) pump, pressurizer level

'ontinued to decrease after the RCDT pump was secured and the drain path isolated.

Operators started a charging pump to recover pressurizer level, but the level reached zero inches prior to being recovered to 20 inches.

The licensee concluded that approximately 1000 gallons of water was added to the reactor coolant system from the refueling water storage tank (RWST) to recover pressurizer level.

Based on the amount of water added, the inspectors determined that the actual pressurizer level did not reach the surge line, and therefore that the pressurizer.was not,emptied.

The licensee later concluded that the pressurizer had not been adequately vented prior to the drain down, which caused a partial vacuum to be drawn in the pressurizer while the RCDT pump was running, and which in turn caused air voids in the control rod

drive mechanisms (CRDMs) and steam generator U-tubes to expand.

When the RCDT pump was secured, the vacuum dissipated and the air voids contracted, causing pressurizer level to continue to decrease.

The licensee concluded that the

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. '-installed.-5/8 inch hose was too small, and should have been at least 3/4 inch to vent the pressurizer.

The licensee generated an ACTION Report (99-0270) to initiate corrective actions.

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Conclusions The licensee made progress in improving system configuration controls, and the training conducted for licensed operators on past configuration control problems was a good initiative. However, several configuration control deficiencies occurred during the current refueling outage which indicate ongoing problems still existed in this area.

The new issues were entered into the licensee's corrective action program; however, this item will remain open pending NRC review of the corrective actions taken for these deficiencies, and a reduction in the number and significance of these events (IFI 50-244/97-10-01).

Nliscelfaneous Operations Issues 08.1 Close L

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ron e Eror L

9-0:" urveila e

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edola io T

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The licensee submitted LER 1999-002 to the NRC on March 29, 1999 in response

.to a February 27, 1999 event when a computer program check surveillance on the axial flux distribution (AFD) monitor was not performed within the required 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> frequency required by ITS surveillance requirement SR 3.2.3.1.

The licensee administratively required that the surveillance be performed every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

However, after a licensed operator successfully performed the surveillance at 3:59 p.m. on February 27, he realized that he could not remember performing it earlier that morning and that it had probably been missed (February 27, 1999 was a Saturday and the operator worked a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift). By consulting the plant process computer alarm printout, the operating crew verified it had not been performed since 11:18 p.m. on February 26, 1999.

The LER concluded that the cause of the event was personnel error by a licensed operator, and indicated that no unusual characteristics existed in the control room at the time the error was made.

The licensee's corrective actions included counseling the operator, emphasizing attention to detail, and having operations supervision review the LER with all operating shifts. The inspectors periodically observed licensed operators performing plant process computer system (PPCS) checks as part of an in-plant evaluation of the LER. The inspectors concluded that the LER adequately described the root cause for this event, and that it identified the necessary corrective actions to prevent recurrence.

Although ITS SR 3.2.3.1 was violated, it represented a

violation of minor significance and is therefore not being cited. This LER is closed (LER 1999-002).

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II. Maintenance M1 Conduct of Maintenance M1.1 General Maintenance Activities a0 Ins ection Sco e (62707)

The inspectors reviewed maintenance work packages and observed portions of plant maintenance activities during the current refueling outage.

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Observations and Findin s The inspectors observed all or portions of the following work activities:

Outage modification installation work (see section M2.2)

Containment recirculation fan cooler heat exchanger coil cleaning Residual heat removal heat exchanger inspections and repairs Routine preventive maintenance on bus 14 normal input breaker Reactor coolant filter replacement The inspectors verified that the correct parts were utilized during the above maintenance, that the applicable industry codes and technical specification requirements were satisfied, that adequate measures were in place to ensure personnel safety and prevent damage to plant structures, systems, and components; and that the operability of the equipment was verified upon completion of post maintenance testing.

The inspector observed overall good quality maintenance during outage work activities, and radiological controls were effective in minimizing workers exposure and controlling contamination.

The licensee identified some deficiencies in circuit breaker operation during maintenance and testing.

The licensee had lowered its threshold for breaker operability during as-found testing, which has resulted in breaker deficiencies being found and corrected prior to being returned to service.

The inspectors noted that the licensee generated ACTION Reports upon the discovery of each deficiency.

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Conclusions The inspectors concluded that the licensee exercised systematic work practices and achieved good quality repairs during preventive maintenance inspections of plant equipment and circuit breakers.

Controlled procedures were in use at maintenance job sites, were up to date and were properly utilized by technicians involved in the outage work. The inspectors observed good personnel and plant safety practices during the maintenance work. A lower threshold for.operability considerations during breaker maint'enance had improved licensee identification and resolution of breaker problem NI1.2 General Surveillance Ac ivities The inspectors observed selected surveillance tests and reviewed the test results for conformance with the specified acceptance criteria for operability.

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Observa ions and Findin s The inspectors observed portions of the following surveillance activities:

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RSSP-24, "Safety Injection Accumulator Check Valve Operability Test." This surveillance accomplished the Sl accumulator discharge test and check valve inservice test (observed March 3, 1999);

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PT-'12.2, "Emergency Diesel Generator B," monthly ITS surveillance for the emergency diesel (observed April 2, 1999);

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RSSP-1.1, "Valve Interlock Verification - RHR System" (observed April 2, 1999)

The inspectors observed that approved test procedures were in use and that procedure details were adequate.

Test instrumentation was properly calibrated and

.used, technical specifications were satisfied, testing was performed by knowledgeable personnel, and test results satisfied acceptance criteria or were properly dispositioned.

The equipment tested met the acceptance criteria specified in the procedures for operability. The licensee placed the test results in the maintenance rule program to trend operability status.

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The inspectors concluded that observed test activities involving safety injection accumulator check valves, emergency diesel generators, and the residual heat removal system were well controlled and the systems were satisfactorily tested to assure operability prior to being returned to service.

M2 Maintenance and Material Condition of Facilities and Equipment M2.1 lns rv'c ins ection Wor i

Pro res The inspectors reviewed the licensee's inservice inspection (ISI) program, and

. observed the performance of non-destructive examination (NDE) and data analysis for plant components inspected during the 1999 refueling outag b.

Observe ions and Findin s The current 1999 refueling outage at the Ginna Station included inservice inspection

- (ISI) activities in.accordance with the American Society. of,Mechanical Engineers (ASME) Code Section XI, as required by 10 CFR 50.55a(g) for the licensee's ISI program.

The 1999 refueling outage was the last outage of the third 10 year ISI interval at Ginna.

An RGSE ISI Engineer maintained the ISI program, compared the ASME Code-required examinations to the planned and completed examinations, and provided for completion of the 10 year interval. The inspectors sampled the ISI program requirements and inspection reports, and noted that the program met ASME Code requirements, or that the licensee had provisions in progress to request relief from the Code for specific cases, as allowed by 10 CFR 50.55a(g).

The inspectors discussed the ISI work planning with cognizant NDE personnel, reviewed portions of the applicable ISI procedures, observed the ISI work in progress, and reviewed documentation of completed work in several ISI areas, The licensee's component inspections for the ASME Code Section XI ISI program included manual and computer-based automated ultrasonic examinations (UT). The inspectors observed the use of a mockup for the UT of the A-and 8-residual heat removal (RHR) heat exchanger lower. head to inlet and outlet nozzles, and a transducer calibration using procedure UT-203, "Manual Ultrasonic Examination of Pressure Retaining Welds in Pressure Vessels from 1/4" to 12"." The mockup provided for the examiners're-inspection practice, and for validation of their

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The mockup also allowed the examiners to optimize their detection techniques using a 60 degree transducer.

The licensee examined the reactor vessel shell ring circumferential welds (no longitudinal welds existed in the shell rings), the inlet/outlet nozzle to shell welds, and the nozzle-to-pipe welds using automated computer-based UT. The licensee calculated the extent of coverage for each weld, and where the extent of coverage was found to be less than 90 percent, RGSE had initiated a provision to request relief from the ASME Code via Relief Request ¹42 for later submittal to the NRC.

The inlet nozzle to shell weld (N28) had an indication that was previously identified in 1979, was confirmed in 1989 to have not grown, and was again examined and reevaluated during the 1999 outage by computer-based UT. The inspector reviewed the licensee's analysis of this indication and noted that it had been extensively documented, was evaluated to have no structural significance, and was acceptable in accordance with the rules of the ASME Code.

In addition to the nondestructive testing and visual inspections designated by the ASME Code, RGSE performed UT examinations of the baffle former bolts located in the reactor vessel lower internals, eddy current test (ECT) examinations of the control rod drive mechanism (CRDM) penetrations through the upper reactor vessel head, UT examination of the four part-length CRDMs, and UT thickness measurements for erosion/corrosion in high energy and service water piping. The licensee also conducted an ECT examinations of all steam generator tubes during the refueling outage.

The inspectors observed the setup and data from UT

examination of baffle former bolts for the reactor vessel internals.

The licensee used a remote ultrasonic, technique to examine these bolts in accordance with 54-ISI-133-00, "Remote Ultrasonic Examination of Baffle Bolts Using Straight Beam Techniques,'~and identified those bolts with crackingin.the head-to-shaft region.

Out of a total of 738 baffle bolts, the licensee completed a successful UT examination on 639, 59 of which revealed indications.

Forty of the bolts with indications were replaced.

The licensee also replaced 16 bolts that did not have indications.

The 19 bolts with indications that remained in the vessel internals were included in an engineering analysis which demonstrated that the final configuration of bolts without indications were acceptable with respect to specific bolt pattern criteria developed by Westinghouse.

The Inconel 600 structural members of the CRDMs that penetrate the reactor vessel head had been shown to be susceptible to stress-related corrosion cracking in the welds on their inside diameter.

Consequently, the licensee conducted ECT examinations of the 37 CRDM penetrations from under the head during the 1999 outage to determine if cracking was present in this area.

Only one CRDM penetration (¹13) had ECT indications in the area of interest, and these were determined to have insignificant depth by subsequent ultrasonic testing.

Prompted by a leak at a similar plant, RG&E also conducted UT examinations of the stainless steel welds (type 304 to type 403) in four partial-length CRDMs on the upper side of the reactor vessel head.

RGRE found no indications in these 'welds using a specialized UT technique.

Prior to and during the 1999 refueling outage, the licensee conducted UT thickness measurements for erosion/corrosion (E/C) in high energy piping, and wall loss or degradation measurements in service water piping under two separate but related inspection programs.

For high energy closed systems, the licensee selected components to be examined based upon the Electric Power Research Institute (EPRI)

"ChecWorks" program, and by inputs from the system engineering, operations, and maintenance departments.

Inspection procedures NDE-602, 603, 604 and 605 provided procedural descriptions and controls over the implementation, grid layout, data processing and UT examination for E/C. The procedures had been recently revised, and a revision was currently in progress for the 1991 Erosion-Control Program Manual.

No issues were identified with the implementation of the current E/C program.

RGSE replaced both steam generators during the 1996 refueling outage, and completed a baseline examination of all steam generator tubes by ECT examination.

During the 1997 refueling outage, all the tubes in both steam generators were also re-inspected by ETC.

No significant degradation was identified at that time. The ECT inspection scope for the steam generator tubes during the current 1999 refueling outage included 50% of the tubes by the bobbin coil technique, and a

...,- sampling of special interest areas including those just.above. the tubesheet, the U-bend support area, and close proximity tubes.

The inspection also included an evaluation of tube buff marks.

The licensee's

"Guidelines for ECT Data Analysis" and procedure ET-109, "Digital Eddy Current Examination of Inconel 690 Steam

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Generator Tubing," provided procedural controls for the ECT process and specified data analysis techniques consistent with current industry practice for steam generator tube examination.

The inspector found the ECT activities to be well

planned and effectively conducted by qualified personnel.

C.

~oc I~sion Inservice inspection activities were well planned and implemented by qualified personnel in accordance with approved procedures.

NRC observation of nondestructive testing activities in progress indicated that the current ISI work was conducted with proper oversight by RGhE staff, and the results were well documented.

The inspections observed were thorough and of sufficient extent to determine the integrity of the components inspected.

Problems identified were properly evaluated and addressed in accordance with ASME Code requirements.

The inspectors identified no significant issues related to the ISI program inspections at Ginna.

Nliscellaneous Maintenance Issues M8.1 Cosed 5 -244

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-01 F rei n a eri lusio F

C tn January 1997, the licensee identified high reactor coolant system (RCS) activity levels, and later determined that failed fuel was the cause.

During the subsequent 1997 refueling outage, the licensee identified one fuel assembly with four pins having breached cladding that was caused by foreign material (metal machining remnants) trapped in the reactor core during the 1996 outage.

During the 1997

~ outage, the licensee also identified several instances of foreign material exclusion (FME) controls that were not adequate to prevent foreign material from becoming-lost, or from entering plant systems (see IR 50-244/97-11).

The inspectors initiated a follow-up item to evaluate the licensee's corrective actions to address the

~ unexpected high incidence of FME problems.

The Maintenance Superintendent was assigned the responsibility to determine the causal factors related to ineffective FME controls and to develop a site-wide effort to implement programmatic improvements.

A multi-disciplinary FME task group was assembled to examine specific work activities where FME had not been adequately controlled, and to identify improved work practices that could reduce the incidence of undesired material from inadvertently entering, or from being left inside plant systems.

The licensee issued interface procedure IP-HSC-1, "Foreign Material Exclusion," to establish general policies and responsibilities at the Ginna Station for preventing the introduction of foreign material into open systems, and to establish formal methods to account for tools and materials used near open systems during maintenance.

The procedure also established and defined foreign material control

.areas and boundaries, with designated postings, barrier.tape, component covers, log books, and material control individuals that were intended to assure that the designated areas were well marked and controlled, and that materials used within the boundaries were properly accounted for. The licensee also issued interface

procedure IP-HSC-2, "System Cleanliness," to define individual responsibilities for maintaining plant system internal cleanliness under three graded levels (Class A, B, and C). The procedure implemented formal controls that were designed to assure

- that minimum cleanliness levels existed following:maintenance on open systems and prior to their operation.

The inspectors reviewed both interface procedures and considered them to be clearly written, with explicit responsibilities and expectations defined for FME controls.

The procedures represented significant programmatic improvements over previous administrative requirements in this area.

The inspectors also directly observed FME areas in place during the current refueling outage, observed the control of materials in and out of these areas, and evaluated the licensee's material accounting methods.

Overall, the licensee's work practices were effective in maintaining a higher level of FME control, and individual plant workers demonstrated a heightened awareness for preventing the inadvertent introduction of foreign material into plant systems.

Material accountability logs around the reactor coolant pumps, the steam generators, the refueling cavity, and the spent fuel pool were well maintained.

In addition, all outage work packages reviewed by the inspectors contained a reference to the appropriate interface procedures, or contained explicit FME controls that were incorporated directly into maintenance work instructions.

During the current outage, the licensee identified several additional occasions where foreign material was inadvertently introduced into plant systems, and initiated ACTION Reports to investigate the causes and to take corrective actions.

A small motor drive gear was lost in a steam generator primary plenum, and several instances where a foreign object fell into the reactor vessel or refueling cavity; for example:

a 15 inch nylon cord was dropped into the cavity during a filter replacement, two "ty-wraps" fell into the reactor vessel during baffle bolt inspections, a small nut and bolt was dropped into the cavity near the reactor vessel, and several paint chips were observed in the cavity and the reactor vessel.

The licensee was able to identify causes for all of the FME incidents that occurred to date during the outage, initiated corrective actions to recover the material, and implemented additional controls to prevent further occurrences, No misplaced foreign material was unaccounted for at the end of the current inspection period.

Based upon the significant programmatic improvements in the FME area, and the licensee's actions to formally incorporate previous weaknesses into their corrective action process, no further need for formal NRC tracking of this issue is necessary.

This item is closed (IFI 50-244/97-11-01).

M8.2 Clos d IFI 0- 4498-07-1: Deficie i

'n Work Pla nin and S r

Pa s

Re ui'i

'n 'Poe i liny n

I P blems This item was opened on July 20, 1998, after an incorrect part was requisitioned for the 9-A reactor trip relay. A 125 VDC relay was required, but a 120 VAC relay was issued from the stock room.

The technician installing the relay noticed the wrong part was obtained prior to its installation.

The planner performing that maintenance activity mistakenly put the 120 VAC relay in the work package, and

the technical reviewer of the package did not identify the error.

The inspectors determined that there was no inventory control problem; however, some aspects of work planning and technical review were insufficient to prevent the improper relay

'from being identified in the work package.

As a result; the licensee generated an ACTION Report (98-0849), and implemented corrective actions including ISC shop training on the lessons learned.

The inspectors determined that the licensee had adequately addressed the deficiency, and noted that no instance has occurred since this event took place that resulted in an incorrect part being installed in the plant.

This item is closed (IFI 60-244/98-07-01).

III. En ineerin E2 Engineering Support of Facilities and Equipment E2.1 Ins alla ion of Plan Modifi ation a0 The inspectors reviewed the preparation and implementation of several plant modifications installed during the refueling outage, and observed portions of the installation work.

b.

erv

'

and Findin i

Ba e

Re c m n s This modification was performed due to aging concerns associated with the existing batteries.

The station batteries had not degraded to the point where replacement was required; however, the licensee concluded that they would have to be replaced prior to the year 2009 and decided to replace them with an upgraded design during the current refueling outage in order to enhance their reliability and to preclude premature failure. The new batteries were a lead/calcium type, where as the previous batteries were lead/acid.

The licensee selected the lead calcium design due to decreased maintenance requirements and longer life. The new batteries also provided 1495 amp-hours, and the previous batteries supplied 1200 amp-hours.

The added size and weight of the batteries required modification and replacement of the existing support racks.

The licensee installed new racks designed to meet the seismic requirements for the battery rooms.

The licensee's safety evaluation for this upgrade supported the conclusions that the modification was within the existing design of the plant.

Com en Coolin Wa er CW Hea Exchan er HX Ins ec i ns and Tube Re lacements During the 1997 refueling outage, the licensee observed higher than expected levels of fouling and pitting corrosion in the A-CCW heat exchanger tubes that were

apparently caused from high siltation and microbiologically induced corrosion (MIC).

At the end of that outage, the licensee had installed a-total of 30,welded plugs in, the B-HX tubes and 45 plugs in the A-HXtubes (each HX contained a total of 1110 tubes).

A subsequent analysis concluded that the number, of plugs would not have reduced the thermal performance of either HX to the extent that they could not accomplish their safety-related function. The licensee conducted ongoing thermal tests during 1998 to determine the actual rate of fouling, and concluded that both HXs would perform adequately throughout the next operating cycle.

However, the licensee decided to completely retube both heat exchangers in the 1999 outage to accommodate improved monitoring and preventive maintenance programs, and to help justify future decisions regarding the optimum long term operation and configuration of both HXs.

The licensee developed Technical Evaluation (TE) 98-0200 to justify the use of new admiralty brass (ASME-SB-111) tubes that were obtained commercially outside the 10 CFR 50, Appendix B procurement program.

Both CCW heat exchangers are safety-related components (Safety Class 3), and a dedication process was required to qualify the commercial material for a safety-related application.

The TE demonstrated that the pressure, temperature, and flow rates would not be changed as a result of the retubing.

Allof the replacement tubes had the identical dimensions for diameter, length, and wall thicknes's'as the original tubes.- Also,:the thermal conductivity of the new tubes was not changed, nor would the physical and hydraulic characteristics of the heat exchangers be changed.

Retubing of the A-and B-HXs was accomplished in accordance with station modification procedures SM-98-200.1 and SM-98-200.2, respectively.

The licensee conducted internal surface inspections for the accumulation of sediment and biological growth (zebra mussels, MIC, etc.).

Both HX tube sheets showed a moderate amount of debris and fouling deposits, with pieces of anode material on their inlet sides, but no substantial degradation of the tubesheets or the HX vessels existed.

The inlet and outlet service water piping showed some evidence of long term internal erosion; however, the licensee considered the condition to be the result of normal operation.

The licensee verified the dimensions of the tubesheet holes prior to installation of new tubes, and then applied a sealant (Locktite 271) between the tubes and tubesheets to compensate for existing mechanical wear in the tubesheets.

The licensee replaced all tubes in each heat exchanger and performed a 100% baseline eddy current test as part of the commercial dedication.

One hole in the tubesheet of the A-HXcould not accommodate a new tube and was plugged.

The testing showed that one tube on the A-HXwas below minimum wall thickness and was replaced.

All remaining serviceable tubes were satisfactory following the eddy current tests.

Quality control (QC) verifications were performed for the final cleanliness (Level C) prior to closing both HXs. The licensee also performed a hydrostatic test of the shell side of both heat exchangers at 165 psig (110% of design pressure)

in accordance with ASME Section IX. A visual examination with QC verifications of the tube/tubesheet joints and all pressure boundary surfaces

showed no leakage.

At the end of the current outage, the licensee continued to evaluate the optimal flow conditions required to prevent excess siltation. (see Section E2.2 below).

c.

Conclusions Plant modifications installed during the current refueling outage were good enhancements to the operation and reliability of plant equipment.

The installation packages reviewed contained detailed instructions and information for performing and documenting the modification work. The safety evaluations reviewed were adequate to demonstrate that the modifications did not represent any unreviewed safety questions.

E2.2 Servic Wa r Valv V-46 0 lns ection and Re la emen The inspectors observed the internal inspection of service water valve V-4620 and reviewed the licensee's analysis of the high throttling conditions that caused internal valve and system degradation, b.

Observe io s and Findin s In December 1998, the licensee discovered a through-wall leak in the A-CCW heat exchanger service water (SW) outlet isolation valve V-4619 that was caused by an unexpected high internal erosion.

Another through-wall leak was found in the SW outlet piping of the B-spent fuel pool (B-SFP) heat exchanger immediately downstream of isolation valve V-8689 that also resulted from high internal erosion.

The licensee concluded that the high erosion in both cases was caused by heavy throttling of the valves and high velocity fluid jets that impinged directly on the areas where the leakage occurred.

Both valves were in close proximity to the auxiliary building SW discharge header, and a relative vacuum in the header contributed to fluid cavitation on the downstream side of both valves that accelerated the erosion.

The licensee modified the piping adjacent to V-4619 to install a smaller valve, and replaced the piping adjacent to V-8689.

UT examinations of the SW discharge isolation valve (V-4620) on the B-CCW heat exchanger did not show internal degradation below the minimum wall thickness.

However, the licensee added the affected SW piping to the erosion/corrosion inspection program and concluded that V-4620 should be opened for internal inspection during the 1999 outage (see IR 50-244/98-13).

Early in the current outage, the licensee closed V-4620 to isolate the B-CCW heat exchanger for retubing (see E2.1 above).

However the valve leaked when fully

. closed, and the valve stem was completely sheared during an. attempt to tightly seat the valve by applying a high torque on its handwheel.

When the valve was later disassembled for internal inspection, the licensee observed significant degradation in areas that were not previously accessible for UT examination.

Some internal

surfaces had experienced a significant loss of base material and deep pitting erosion.

The licensee considered repairing V&620 in place; however, its overall condition warranted a complete replacement.

Therefore the licensee developed a

plant change request (PCR 99-020) and issued a modification package to replace V-4620 with a smaller valve identical to V-4619. Both V-4619 and V-4620 were originally 'l4 inch plug valves, and were replaced with 10 inch stainless steel butterfly valves similar to V-8689. The licensee considered that the replacement valves would be less susceptible to erosion white throttled under the existing system conditions, as had been previously observed in V-8689. The inspector reviewed the PCR 99-020 package and determined that the modification was well supported with a stress analysis under criteria specified in'the original design code (B31.1) for the SW piping. The safety evaluation included in the modification package adequately demonstrated that the modified valve and piping remained within the existing design basis of the SW system.

,C Each CCW heat exchanger was designed with sufficient thermal capacity to remove 100% of the accident heat loads in the CCW system, and the licensee had historically operated with both heat exchangers in service under normal plant power conditions.

This required heavy throttling (5-15% open) of both V-4619 and V-4620 to prevent over cooling of the CCW system during normal operation.

'However, the licensee concluded that the reduced SW flow was responsible for a high fouling rate in the A-CCW heat exchanger and the high erosion at both SW discharge valves.

The licensee initiated ACTION Report 97-2149 and CATS item N)07592 to determine the optimal SW alignment and flowto both CCW heat exchangers.

After the V-4620 modification and CCW heat exchanger maintenance were completed, both V-4619 and V-4620 remained heavily throttled. At the end of the inspection period, engineering and operations were evaluating several options for reconfiguring the CCW and SW systems to balance the need to reduce siltation and erosion rates, and the need to optimize CCW system temperatures.

The licensee planned to complete these actions by the end of the current refueling outage.

Conclusions The licensee successfully resolved internal valve degradation that resulted from heavy throttling of service water valve Y-4620 by replacing it with a smaller valve that was more resistant to erosion.

However, the system conditions that required heavy throttling of service water at the CCW heat exchangers were not yet resolved.

The licensee continued to evaluate the need to increase service water flow to reduce siltation and erosion, and to maintain optimal CCW system temperature ~

W

E8 Miscellaneous Engineering Issues E8.1 Closed IFI 50-244 98-012-03 Service Wa er Valv Testin Procedure RSSP-2A.4, "CNMT Recirculation Fan Service Water Valves Leak Check," contained a provision to cycle the inlet butterfly valves necessary to isolate the Recirculation Fan Cooler Valves for testing as part of the ASME Section XI inservice testing (IST) program.

RGRE agreed to review this procedure,to determine if it represented a preconditioning issue.

Part 3.5.11 of the Inservice Testing Program (procedure IP-IIT-2) discussed preconditioning of components.

It presented the background and controls to avoid preconditioning.

The provision in RSSP-2.4A to cycle the inlet butterfly valves was applicable to valves that were not being tested and were not part of the ASME Section XI IST program, but were necessary to be leak-tight to create the necessary conditions for testing of designated ASME Section XI valves.

No needed changes in procedures, training or guidance to avoid preconditioning were identified by RGSE or NRC. This IFI is closed (IFI 50-244/98-12-03).

E8,2

e i e see v n e o LER 1999-

"Def'

n ies'S V

dor S eamli e Break Mas a

E r

le se A al si esul i

a Bei Outside The licensee submitted LER 1999-001 to the NRC on March 25, 1999, in response to a notification from its nuclear steam supply system (NSSS) vendor (Westinghouse)

on February 22, 1999, of two modeling errors in the plant's 1995 analysis for a main steamline break (MSLB) inside containment with an assumed single failure of a main feedwater regulating valve (FRV). One error involved a large volume of high temperature feedwater that was not previously accounted for; i,e.,

approximately 1000 gallons upstream of the failed FRV was not included in the existing analysis.

The other error involved an under estimate of the total time

, required to isolate the feedwater system; i.e., the main feedwater pump isolation valve actually took 80 seconds to close, and the existing analysis assumed approximately 15 seconds.

Both modeling errors resulted in a nonconservative impact on the actual margin to the design basis maximum containment pressure (60 psig), and preliminary analyses indicated that it could have been exceeded following the more limiting MSLB scenario.

The licensee reported that the nonconservative errors in the 1995 MSLB analysis were also reportable under 10 CFR 21.

At the time of the notice, the Ginna plant was at approximately 75% power and was undergoing a gradual power reduction for a planned refueling outage that commenced on March 1, 1999.

The licensee initiallyperformed an analysis with assumptions that accounted for a cycle-specific shutdown margin and integral flow restrictors in the new steam generators, and that credited the current winter-time heat sink (service water) temperatures.

As a result, the licensee placed an operating restriction on the maximum service water temperature of 40 degrees Fahrenheit ('F)

tg to assure that adequate margin existed in the heat removal capacity of th'e containment recirculation fan coolers.

The actual service water temperature at, the time was approximately 35'F.

The licensee reported in the LER that there were no operational or safety consequences to the Ginna plant at the time of the event since the preliminary analysis showed that the maximum containment pressure would not be exceeded if a MSLB occurred while the power reduction continued, and if service water remained less than 40'F.

The inspectors reviewed the preliminary analysis with engineering and licensing personnel involved in it's development, and considered it's assumptions reasonable to demonstrate that the plant would remain within its design basis for the remainder of the cycle. -By March 5, 1999, the Ginna plant was shutdown in MODE 5, and the temporary operating restrictions no longer applied.

The LER indicated that the root cause of the issue was the modeling errors in the 1995 MSLB analysis.

However, the LER did not indicate how the modeling errors occurred or why RGSE's normal design verification processes did not detect them.

LER 1999-001 stated that further analysis was necessary to determine ifthe plant was actually operated outside its design basis during previous operating cycles at high power, and to assure that the maximum containment pressure would not be exceeded if a limiting MSLB occurred during the next operating cycle. At the end of the inspection period, RGSE and Westinghouse were preparing a safety evaluation for the cycle 28 reload that would address the major mass-energy release assumptions in the limiting MSLB at Ginna and the resulting peak containment pressure.

The licensee indicated that redefining some overly conservative assumptions in the 1995 analysis (e.g., instantaneous flashing of the feedwater into containment) could recover most of the desired margin to maximum containment pressure, and that the results would be reported to the NRC in a future supplement to the LER. Meanwhile, the pending reload analysis represented a MODE 3 operating restriction and prevented a plant restart before the licensee could review and approve it. This LER remains open pending NRC review of the specific root cause of the 1995 modeling errors, the cycle 28 reload analysis, the LER supplement, and an evaluation of deficiencies in RGRE's design verification processes that did not detect the errors in the 1995 analysis (LER 1999-001).

IV. Plant Su o

R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 lm e

nt

'o f h Radia ion Pro e ion Pro ram a0

~..

From March 15 through March 19, the inspectors reviewed the licensee's implementation of their "As Low As Reasonably Achievable" (ALARA)program and the "Ginna Refuel Outage ALARAPlan," relative to work planning and control in support of the refueling outage for the period of March 15 through March 19. The

review included an evaluation of performance related to job-specific ALARA reviews, radiological control records, interviews with staff and selected workers, and direct observation of radiological controls established for the in-service inspection of reactor components.

ll The inspectors also evaluated the licensee's performance relative to applicable requirements of 10 CFR 20, and relevant licensee procedures related to ALARAjob review preparation, radiological surveys, and radiation area postings.

Observations and i din s The licensee's overall planning and preparations to minimize radiation dose to personnel, and to limitthe spread of contamination during the inservice inspection of reactor components and outage-related activities, were comprehensive.

The "Ginna Refuel Outage ALARAPlan" detailed the prerequisites for the project requiring specific radiological controls, incorporated lessons learned from experience gained at other facilities, and identified individual staff responsibilities to assure that emergent ALARAissues were promptly addressed in the work control process.

The licensee effectively used system flushing, remote cameras, robotic equipment, and temporary shielding to limit personnel exposure.

The ALARAgroup provided detailed pre-job briefings and monitored the effectiveness of dose control measures; and routinely communicated their assessments to site management through in-progress ALARAreviews and status reports.

In response to assessments by the ALARAgroup, additional operational and engineering controls were effectively used to further reduce the dose rates in various work areas.

These timely measures further reduced the general area dose rates and resulted in less cumulative exposure.

Pre-job ALARAand radiation work permit (RWP) briefings for RHR heat exchanger testing comprehensively detailed the radiological conditions at the job site, dosimetry/protective clothing requirements, low dose waiting areas, work expectations, and contingency actions.

Workers were knowledgeable of radiological conditions and implemented the requisite controls.

Radiation Protection (RP)

supervision and technicians were actively involved in overseeing that exposure controls were properly used, and that emergent issues were promptly addressed.

This was exemplified by the prompt initiation of an ACTION report following identification of low level foot contamination found on workers in the auxiliary building.

In response to the subsequent evaluation, and the licensee effectively carried out additional measures to limitgeneration of contamination during testing of the RHR heat exchangers.

No unplanned exposures or significant contamination events occurred during the outage.

Cumulative personnel exposures were maintained below projected estimates,

S

c.

Conclusions r

!

The licensee's ALARAprogram requirements were well developed, integrated in the work control process and effectively implemented during the in-service inspection of reactor components.

Radiation doses to individuals and work groups were closely monitored by the ALARAgroup and provided to management for timely resolution of emergent issues, resulting in cumulative doses below the estimated levels.

R2 Status of RP&C Facilities and Equipment R2.1 a

i lo ical W r Pr ic s Access Control and H usekee in At various times the inspectors accompanied RP Section management and technicians, and independently toured site areas including the containment building, auxiliary building, and intermediate building to observe radiological practices, access controls, and housekeeping.

The inspectors evaluated the licensee's performance relative to the requirements contained in 10 CFR 20 and applicable licensee procedures.

b.

Ob v

io Fin in

~ RMIPs were complete and accurate with current survey data.

Radiological surveys were comprehensive and accurately characterized the radiological conditions in the work areas.

Laborers and technicians were knowledgeable of radiological conditions and complied with RWP requirements.

Radiologically controlled areas (RCAs) were properly posted and access appropriately controlled.

Contamination control measures were conscientiously implemented at job sites visited by the inspectors.

Locked high radiation areas located in the auxiliary building were properly posted and their keys were properly controlled.

The licensee appropriately performed daily source checks of survey instruments and portal instrumentation, and adequately controlled the issuance of radiation survey instruments.

Survey instrument calibration records were current and complete, and sealed source inventories and leakage tests were performed within the required frequency.

The respiratory protection program was appropriately administered.

Personnel respirators were properly maintained, and medical qualifications and fittest data for selected workers were current and readily retrievable.

Airsampling equipment was appropriately calibrated and sampling points provided representative data.

The inspectors interviewed selected workers, reviewed training records, and interviewed workers, and determined that technicians were knowledgeable of current radiological controls and job-specific requirements.

Contractor radiation technicians were appropriately screened, trained, and qualified to carry out their

responsibilities.

Shift turnovers between radiation protection supervision were detailed, and job status and emerging issues were thoroughly discussed..

.,

~. - -Personal dosimetry was appropriately worn in radiologically controlled areas.

Extremity dosimeters were used when appropriate, and whole-body dosimeters were repositioned for specific jobs to assure that the maximum exposure was monitored.

Dosimetry records were current and properly maintained.

Internal dosimetry was conservatively performed, and Derived AirConcentration-hours (DAC-hrs) for workers were appropriately tracked.

Housekeeping was generally satisfactory with only isolated discrepancies identified.

Walkways were unobstructed, potentially contaminated waste materials were segregated from clean materials, and work area contamination control boundaries were in place.

c.

C ncl ions The radiological controls program was effectively implemented as evidenced by a qualified and experienced staff that properly implemented procedures to minimize external and internal exposure.

The licensee developed detailed radiation work permits, appropriately monitored personnel exposure, and adequately maintained radiologically controlled areas.

R7 Quality Assurance In RP&C Activities R7.1 A di urveillan es an Man em n A ralsals o e (83760)

The inspectors reviewed a sample of the licensee's audits, surveillance reports, and various management appraisals to determine the adequacy of identifying, evaluating, and correcting deficiencies related to implementation of the radiation protection program.

b.

Observa ions and findin s The Respiratory Protection Audit (AINT-1998-0021-JMT), the Radiation Protection and Chemistry Audit (AINT-1998-0007-TGT), and the Radwaste/Process Control Program Audit (AlNT-1998-0014-JMT) were in-depth program assessments that balanced observations of worker practices with procedure reviews and verification of regulatory compliance.

The licensee appropriately addressed Abnormal Condition Tracking Initiation or Notification (ACTION) Reports initiated from audit findings, and appropriately addressed and closed them in a timely manner.

A sample of routine QA surveillance reports conducted in 1999 evaluated specific aspects of the radiation protection program, including the work planning process, contamination control measures, adequacy of pre-job briefings, and adherence to

t RWP requirements.

During the outage, surveillances were conducted weekly, during normal hours and on the back-shift, and in all plant radiologically controlled areas.

Routine plant.tours.and observations of work-in-progress by the Radiation Protection Section management and staff were required by site procedures and conscientiously carried out. Attributes addressing human performance and the effectiveness of work planning were rated, and suggestions for improvement were described.

ACTION Reports were initiated at a conservative threshold to address off-normal conditions or trends.

Several ACTION Reports were reviewed including No.

99-0365 {potential unmonitored release path through containment equipment hatch), No. 99-0324 (workers observed to be removing protective clothing in wrong sequence),

and No. 99-0350 (personnel hot particle contamination).

These problems were appropriately evaluated and resolved in a timely manner with reasonable corrective actions.

onclu i n Work performance standards were effectively monitored and reinforced by close and frequent QA and management oversight.

Off-normal conditions were conservatively identified, appropriately evaluated, and resolved in a timely manner.

Miscellaneous ROC Issues os d

I 5 - 44 97-1-

C vi fueli The licensee addressed this issue through their Commitment and Action Tracking System (CATS No. RO684200).

Prior to the current inspection, a Technical Support Request (TSR 021257) had been initiated to evaluate the source of the leakage, and to determine the necessary actions to control or eliminate it. During the current inspection, an engineering evaluation was in progress (Work Order 19802436) to determine the nature and extent of the condition.

White the cavity and transfer canal were flooded for refueling operations, the licensee investigated the leakage and identified that a leak path existed through the cavity liner above a four foot elevation inside the containment.

The licensee utilized compensatory measures and assigned additional personnel to keep the leaking water from accumulating in the

~

containment basement during the current outage.

The inspectors observed the leakage and considered the licensee's current controls to be adequate.

The lead engineer investigating the leakage continued to evaluate possible solutions, which included methods to precisely locate the leak path in the liner, and to seal the canal in a similar manner previously used to stop transfer canal leakage from the auxiliary building into the residual heat removal pump room (see IR 50-244/97-11).

This issue was appropriately addressed by the licensee and could be resolved within their corrective action system, Formal tracking by the NRC is no longer required and this item is closed (IFI 60-244/97-11-04).

S1 Conduct of Security and Safeguards Activities S1

~ 1 Securit and Safe uards a.

Ins ection Sco e (81700)

During March 15 - 19, the inspectors evaluated the licensee's security program to determine whether security and safeguards activities met the licensee's commitments in the NRC-approved security plan ("the Plan" ) and met NRC regulatory requirements.

Specific areas inspected included the access authorization (AA) program; operation of alarm stations; communications; and protected area (PA)

access control of personnel, packages and vehicles.

b.

Obs rva io and Findin cc s horiza ion AA Pro ram: The AA program was reviewed to verify implementation was in accordance with applicable regulatory requirements and Plan commitments.

The review included an evaluation of the effectiveness of the AA procedures, as implemented, and an examination of AA records for 6 individuals.

Records reviewed included both persons who had been granted and had been denied access.

The AAprogram, as implemented, provided assurance that persons granted unescorted access did not constitute an unreasonable risk to the health and safety of the public. Additionally, access denial records and applicable procedures were

- reviewed to verify that appropriate actions were taken when individuals were denied access or had their access terminated.

(CAS) and the secondary alarm station (SAS).

Both alarm stations were equipped with appropriate alarms, surveillance and communications capabilities.

Inspector interviews with alarm station operators found them knowledgeable of their duties

-. and responsibilities.

Inspector observations and interviews also verified that the alarm stations were continuously manned, independent, and diverse so that no single act could remove the licensee's capability to detect a threat and call for assistance.

The alarm stations did not contain any operational activities that could interfere with the detection, assessment, and response functions.

with alarm station operators, and determined that the alarm stations were capable of maintaining continuous communications with each security force member (SFM)

on duty. Alarm station operators had tested communication capabilities with the local law enforcement agencies as committed to in the Plan.

Protec ed Area PA Access Control of Person el nd Hand-Carried Packa es:

During peak activity periods on March 16 and 17, inspectors observed personnel and package search activities at the personnel access portal.

Positive controls were in place to ensure only authorized individuals were granted access to the PA, and that all personnel and hand-carried items entering the PA were properly searche Protected Area PA Access Control of Vehicles:

On March 18, the inspectors observed vehicle search activities at the vehicle access gate.

The vehicle search was thorough and accomplished in accordance with commitments in the Plan.

The active land vehicle barrier was also operated in accordance with Plan commitments.

c.

Conclusions The licensee was conducting its security and safeguards activities in a manner that protected public health and safety.

As implemented, the existing program for access authorization, access control, and alarm station operation met the licensee's commitments and NRC requirements.

S2 Status of Security Facilities and Equipment S2.'I Securi Fa ili i s and E ui men The inspectors evaluated the licensee's PA assessment aids; the PA detection aids, and personnel search equipment.

b.

bserva ion a 'i of the licensee's assessment aids in the Central Alarm Station (CAS), at the Secondary Alarm Station (SAS), and by observing the PA perimeter.

The assessment aids were of good quality and provided zone overlap.

To ensure the Plan's commitments were satisfied, the licensee also had procedures in place requiring the implementation of compensatory measures in the event the alarm station operator was unable to properly assess the cause of an alarm.

• "

detection zones in the plant's protected area.

The appropriate alarm was generated in each zone for each test.

Through observations and review of the testing documentation associated with the equipment repairs, the inspectors verified that repairs were made in a timely manner, and that the equipment functioned effectively, and met the pertinent commitments in the Plan, Pers nnel and Pa ear h E ui me:

On March 16, inspectors observed both the routine use and the daily operational testing of the licensee's personnel and package search equipment.

Personnel search equipment was tested and maintained in accordance with licensee procedures and the Plan, and personnel and packages were properly searched prior to PA access.

Inspector observations and procedural reviews determined that the search equipment were performed in accordance with licensee procedures and Plan commitment !

c.

Conclusions

The licensee's security facilities and equipment were well maintained and reliable, and were able to meet the licensee's commitments and NRC requirements.

S3 Security and Safeguards Procedures and Documentation S3.1 Securit Pro ram Procedures The inspectors reviewed and evaluated the licensee's security program implementing procedures and security event logs.

b.

bs

'ons a

'n i

Securi P o ram ced r s:

Inspector review of selected security program implementing procedures verified that the procedures were consistent with the Plan's commitments.

Se u 've t Lo s: The inspectors reviewed the security event logs for the previous twelve months.

Based on this review, and on discussions with security management, the inspectors determined that the licensee had appropriately

,analyzed, tracked, resolved and documented safeguards events that they determined did not require a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> report to the NRC.

The licensee properly implemented security and safeguards procedures and properly maintained program documentation.

Event Logs were also properly maintained and effectively used to analyze, track, and resolve safeguards events.

S4 Security and Safeguards Staff Knowledge and Performance S4.1 S aff Knowled e and Performance The inspectors evaluated the licensee's security staff's requisite knowledge.

b.

bserv ions an Securi Force Re isi e Knowled e: The inspectors observed a number of SFMs during the performance of their routine duties.

These observations included alarm station operations, personnel and package searches, and exterior patrol alarm response.

The inspectors interviewed several SFMs and determined that they were

knowledgeable of their responsibilities and duties, and could effectively carry out their assignments.

c.

Conclusions The SFMs adequately demonstrated that they had the requisite knowledge necessary to effectively implement the duties and responsibilities associated with their position.

S5 Security and Safeguards Staff Training and Qualifications (T&Q)

S5.1 Staff Trainin and Qualifica ions aO lns ection Sco e (81700)

The inspectors evaluated security personnel training and qualifications, and the maintenance of security training records.

b.

Observa ions and Findin s

Se uri Trainin and Qualifica io: On March 17, the inspectors reviewed the licensee's T&Q records for ten SFMs.

The results of the review indicated that these personnel were trained in accordance with the licensee's approved T&Q plan.

Jt:

i h>>,

h they were properly maintained, accurate, and reflected the current qualifications of the SFMs.

C.

C nclusion Security force personnel had been trained in accordance with the requirements of the licensee's training and qualification plan. Training documentation was properly maintained and accurate and the training provided by the security training staff was effective.

S6 Security Organization and Administration S6.1 a

a en Su o

d ff civ ess hHLKLi~I The inspectors evaluated security and safeguards management support, management effectiveness, and security staffing level b.

Observa ions and Findin s Mana ement Su ort: The inspectors'eview of program implementation since the last program inspection disclosed that adequate resources.and support continued to be available to ensure effective program implementation.

Mana ement Ef ectiveness:

The inspectors reviewed the security management organizational structure and reporting chain, and noted that the Manager of Security's position in the organization provided a means to make senior RGSE management aware of programmatic needs.

met the requirements specified in the Plan and implementing procedures.

C.

The level of management support was adequate to ensure effective implementation of the security program, and was evidenced by the allocation of resources to support programmatic needs.

S7 Quality Assurance (QA) in Security and Safeguards Activities S7.1 iv ss f QAA iviie ao The inspectors evaluated the licensee's security audits, problem analyses, corrective actions, and the effectiveness of management controls.

b.

erv i n a

Fi din

~d~:

The inspectors reviewed the licensee's August 1998 QA Security Program audit (AINT-1998-0011-TGT), and their 1998 Fitness-for-Duty audit (AINT-1998-0012-TGT). The audit checklists disclosed that the audits included all components of the security program and were comprehensive in scope.

Both audit teams included an independent technical specialist, and were conducted in accordance with regulatory requirements.

Findings from the audits were not indicative of program weaknesses, and implementation of corrective actions for the findings were generally to affect program enhancements.

i f

program indicated that potential weaknesses were being properly identified, tracked, and trended.

Corrective Actions: The inspector reviewed corrective actions implemented by the licensee in response to the QA audits and self-assessment program, and determined that the corrective actions were effectiv Effectivene s of Mana ement Con rois: The inspector observed that the licensee had programs in place for identifying, analyzing and resolving security issues.

They included the performance of annual QA audits, a departmental self-assessment program, and the use of industry data, such as violations of regulatory requirements identified by the NRC at other facilities as criteria for self-assessments.

C.

Conclusions Inspector review of the licensee's audit program indicated that they were comprehensive in scope and depth, that the audit findings were reported to the appropriate level of management, and that the program was being properly administered, In addition, inspector review of the documentation applicable to the self-assessment program indicated that the program was being effectively implemented to identify and resolve potential weaknesses.

V. IVIana emen Meeti s

XO INPO Plant Evaluation Report Review On April 2, 1999, the inspectors reviewed the preliminary results of the recent Institute for Nuclear Power Operations {INPO) plant evaluation conducted from February 1 through February 12, 1999.

The inspectors determined that the findings were consistent with those identified by the NRC.

X1 Exit Meeting Summary The inspectors for the inservice, physical security, and radiological control inspections met with licensee representatives at the conclusion of those inspections on March 19, 1999. Atthat time, the purpose and scope of the inspections were reviewed, and the preliminary findings were presented.

The licensee acknowledged the preliminary inspection findings.

After the inspection period was concluded, the inspectors presented the overall results to members of licensee management on April 9, 1999.

The licensee acknowledged the findings presented.

The licensee considered that no proprietary information was included in the inspection result ATTACHMENTI PARTIALLIST OF PERSONS CONTACTED Licensee G. Graus G. Hermes J. Hotchkiss G. Joss E. Palmer J. Pascher R. Ploof P. Polfleit R. Popp J. Smith W. Thomson J. Widay T. White G. Wrobel Electrical/ISC Maintenance Manager Acting Primary Systems Engineering Manager Mechanical Maintenance Manager Results and Test Supervisor Security Supervisor Electrical/ISC Systems Engineering Manager Secondary Systems Engineering Manager Emergency Preparedness Manager Production Superintendent Maintenance Superintendent Radiological Protection 5 Chemistry Manager Plant Manager

, Operations Manager Nuclear Safety & Licensing Manager INSPECTION PROCEDURES USED IP 37551:

IP 40500:

IP 61726:

IP 62707:

IP 71707:

IP 71750:

IP 81700:

IP 83750 Onsite Engineering Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems Surveillance Observation Maintenance Observation Plant Operations Plant Support Physical Security Program for Power Reactors Occupational Radiation Exposure IP 92700:

IP 92901:

IP 92902:

IP 92903:

Onsite Follow-up of Written Reports of Nonroutine Events at Power Reactor Facilities Follow-up - Operations Follow-up - Maintenance Follow-up - Engineering

Attachment I

ITEMS OPENED, CLOSED, AND DISCUSSED

~Oened LER 1999-001:

Deficiencies in NSSS Vendor Steamline Break Mass and Energy Release Analysis Results in Plant Being Outside its Design Basis Closed LER 1999-002:

Surveillance Not Performed, Due to Personnel Error, Resulted in Violation of Technical Specifications IFI 98-12-03:

IFI 98-07-01:

Service water valve testing Deficiencies in Work Planning and Spare Parts Requisitioning; Potential

. Inventory Control Problems IFI 97-1 1-04:

IFI 97-11-01:

Reactor Cavity/Refueling Canal Water Leakage Foreign Material Exclusion (FME) Controls

~Dl~sse l

IFI 97-10-01:

Weak Configuration Control LIST OF ACRONYMS USED AA AFD ALARA ARV ASME CAS CATS CCW CFR CNMT CRDM DAC E/C ECT ESF ft-Ibs FME access authorization axial flux distribution As Low As Reasonably Achievable atmospheric relief valve American Society of Mechanical Engineers central alarm system Commitment and Action Tracking System component cooling water Code of Federal Regulations containment control rod drive mechanism Derived Air Concentration erosion/corrosion eddy current test engineered safety feature foot-pounds foreign material exclusion

Attachment I

FRV IFI IR ISI IST ITS LCO LER MIC MSLB NRC NDE PA PPCS PT pslg QA QC RCA RCDT RGSE RHR RP RPtkC RTD RWP RWST SAS SFM SI Tave TRQ TE TSR UT VAC VDC VT feedwater regulating valve inspector follow-up item inspection report

- inservice inspection inservice test Improved Technical Specification limiting condition for operation Licensee Event Report microbiologically induced corrosion main steam line break Nuclear Regulatory Commission non-destructive, examination protected area plant process computer system periodic test pounds per square inch gage quality assurance quality control radiologically controlled area reactor coolant drain tank Rochester Gas and Electric Corporation residual heat removal Radiation Protection Radiological Protection and Chemistry resistance temperature detector radiation work permit refueling water storage tank secondary alarm system security force member safety injection primary coolant average temperature Training and Qualification technical evaluation technical support request ultrasonic test volts alternating current volts direct current visual test

e