IR 05000244/1996012

Insp Rept 50-244/96-12 on 961116-970104.No Violations Noted. Major Areas Inspected:Licensee Operations,Engineering,Maint & Plant Support

ML17264A819
Person / Time
Site:	Ginna
Issue date:	02/07/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML17264A818	List: IR 05000244/1996012 ML17264A817
References
50-244-96-12, NUDOCS 9702130253
Download: ML17264A819 (31)

Text

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

License No.

DPR-18 Report No.

50-244/96-1 2 Docket No.

50-224 Licensee:

Facility Name:

Location:

Rochester Gas and Electric Corporation (RG&E)

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

November 16, 1996 through January 4, 1997 Inspectors:

P. D. Drysdale, Senior Resident Inspector C. C. Osterholtz, Resident Inspector Approved by:

Lawrence T. Doerflein, Chief Projects Branch

Division of Reactor Projects 9702130253 970207 PDR ADOCK 05000244

PDR

EXECUTIVE SUNIMARY R. E. Ginna Nuclear Power Plant NRC Inspection Report 50-244/96-12 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support.

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

~Oe rations The licensee's operability assessment in response to periodic spikes in average reactor coolant system temperature (Tave) adequately addressed Updated Final Safety Analysis Report (UFSAR) accident analysis concerns regarding over power delta T and over temperature delta T reactor trip setpoints, and verified that the temperature spikes did not result from an instrument malfunction.

The licensee was continuing to evaluate the impact of slight, but inadvertent reactivity additions, and the frequent and unnecessarily cycling of the control rods to contribute to premature system failures.

Control room operators and the fire brigade responded appropriately to an automatic trip of the C-Service Water (SW) pump.

Service water system flow was returned to normal immediately, and the cause of the screenhouse fire alarm was investigated and identified within five minutes of annunciation.

The apparent cause of the fire alarm was smoke from a failed motor due to a winding short.

Maintenance Overall, maintenance and surveillance activities were performed effectively and were well coordinated with all site organizations.

Maintenance technicians demonstrated good technical abilities and performed their work in accordance with procedure requirements.

Maintenance personnel responded appropriately to a testing failure of the B-SW pump breaker, and successfully identified and corrected the deficiency in the breaker.

The licensee is evaluating potential common failure mechanisms for similar breakers in the plant.

Maintenance and engineering personnel worked well together in troubleshooting a testing failure of the B-Emergency Diesel Generator (EDG) breaker, and in a short time identified and resolved the equipment problem.

The licensee's plan to perform another periodic test (PT) with diagnostic test equipment was appropriate since the breaker has had recent test failures.

Instrumentation and Control (IKC) personnel successfully identified and resolved the operability problems with the steam dump valve controller.

Additionally, the briefing conducted between IRC and operations personnel was thorough and demonstrated good coordination and communication between the two department Executive Summary (cont'd)

The licensee took appropriate action in response to a noted increase in the charging pump leak rate.

The corrective actions were successful in that the measured leak rate dropped significantly following the B-charging pump packing replacement.

The licensee's efforts to reduce charging system piping vibrations had reduced the magnitude of the problem, but the root cause had not yet been fully identified.

~En ineerin The engineering department's identification that the SW pump control power transfer switches and local start buttons could fail from a collapse of the intermediate building block wall was a good finding. The modification adequately resolved the concern for a potential loss of all service water should the block wall fail, and the safety evaluation performed appropriately focused on accident and fire protection issues.

The licensee's efforts in identifying and evaluating the potential for overpressurization of the containment charcoal filter dousing header were both effective and timely. The final resolution to install a relief valve in the line was appropriate, in that the functionality of the containment spray system was maintained without eliminating the charcoal filter dousing capability.

The licensee's ongoing efforts to evaluate the use of the area radiation monitor (R-5)

installed near the spent fuel pool as a criticality monitor for the new fuel preparation area (NFPA) were well focused on meeting the specifications of 10 CFR 70.24.

However, pending the outcome of the licensee's evaluation and additional review by the NRC, this item is unresolved (URI 60-244/96-12-01).

Plant Su ort The licensee apparently resolved the contamination boundary control and radiological labeling deficiencies in the auxiliary building.

The licensee's actions were not programmatic improvements and the inspectors questioned whether they would prevent further recurrence TABLE OF CONTENTS EXECUTIVE SUMMARY

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Il TABLE OF CONTENTS...,

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iv I. Operations

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08 Conduct of Operations

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01.1 General Comments 01.2 Summary of Plant Status

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Operational Status of Facilities and Equipment

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02.1 Average Reactor Coolant System Temperature (Tave) Swings

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Operator Knowledge and Performance......

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04.1 Operator Response to a Service Water (SW) Pump Trip.....

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Miscellaneous Operations Issues (92700)

08.1 (Open) LER 96-013: Circuit Breakers Closed While in Mode 3, Due to Personnel Error, Resulted in Condition Prohibited by Technical Specifications.

08.2 (Closed) LER 96-014: Pressure Relieving Capability Could be Degraded Due to Single Failure of DC Power, Which Could Prevent Mitigating the Consequences of an Accident........

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II. Maintenance

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M1 Conduct of Maintenance

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M1,1 General Comments on Maintenance Activities M1.2 General Comments on Surveillance Activities.....

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M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Failure of the B-SW Pump Breaker During Post-Maintenance Testing

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M2.2 B-Emergency Diesel Generator (B-EDG) Output Breaker Troubleshooting M2.3 Steam Dump Valve Pressure Controller Replacement M2.4 B-Charging Pump Packing Leakage and Charging System Piping Vibrations

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

E2.1 Service Water Pump Control Power Transfer Switch Modification

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E2.2 Potential Overpressurization of the Containment Charcoal Dousing Header

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E2.3 Criticality Monitor For The New Fuel Preparation Area...

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1 3 R1 Radiological Protection and Chemistry (RPtkC) Controls............

R1.1 Contamination Boundary Control and Radiological Labeling....

V. Management Meetings....

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Table of Contents (cont'd)

X1 Exit Meeting Summary L1 Review of UFSAR Commitments

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ATTACHMENT Attachment 1 - Partial List of Persons Contacted Inspection Procedures Used Items Opened and Closed List of Acronyms

Re ort Details I. 0 erations

Conduct of Operations'1.1 General Comments Ins ection Procedure IP 71707 The inspectors observed plant operation to verify that the facility was operated safely and in accordance with licensee procedures and regulatory requirements.

This review included tours of the accessible areas of the facility, verification of engineered safeguards features (ESF) system operability, verification 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 for out-of-service equipment were implemented, and verification that logs and records accurately identified equipment status or deficiencies.

01.2 Summar of Plant Status The p1ant operated at full power (approximately 100/o) throughout the inspection period.

On December 25, 1996, the C-service water (C-SW) pump tripped due to a phase-to-phase short in its motor windings.

Control room operators appropriately responded by manually starting the D-SW pump.

There were no other significant operational events or challenges during the inspection period.

Operational Status of Facilities and Equipment 02.1 Avera e Reactor Coolant S stem Tem erature Tave Swin s a.

Ins ection Sco e (71707)

The inspectors reviewed the vperability assessment performed in response to occasional swings in Tave.

b.

Observations and Findin s Following the plant startup on November 12, 1996, control room operators noted occasional spikes on temperature instrument Tl-402 (Tave channel 2), and on TI-406 (delta T channel 2). The temperature changes appeared to be in the magnitude of approximately 2 degrees fahrenheit (

F) for delta T and 1'F for Tave.

The spikes were in the decreasing direction and lasted between 10-15 seconds.

The temperature spikes on Tave channel 2 were also initiating automatic control rod movement in the "rods out" direction, averaging about one step every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

This movement resulted in slight inadvertent positive reactivity additions to the

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

Individual reports are not expected to address all outline topic reactor.

After each occurrence, control room operators would manually insert the control rods one step in response to the reactivity addition.

I< was not clear whether the indicated temperature changes were the result of an instrument malfunction, or actual coolant temperature variations.

The inspectors were concerned that inadvertent reactivity additions, even though slight, occurred in the plant.

The inspectors were also concerned that the rod control system was being unnecessarily cycled, potentially increasing the probability of a system failure.

The licensee initiated an ACTION Report (96-1142) to address the anomaly, which included an operability assessment.

The assessment concluded that the indication changes were actual temperature variations caused by coolant vortexing in the reactor coolant system (RCS), combined with the location of the hot leg resistance temperature detectors (RTDs) for channel 2. Channel 2 utilized two hot leg RTDs in approximately the same location in the A-RCS loop.

The assessment also concluded that sufficient margin was maintained for the reactor trip setpoints for over power delta T and over temperature delta T, based on the accident analysis and reactor trip setpoint criteria established in the Updated Final Safety Analysis Report (UFSAR) and the Improved Technical Specifications (ITS). Additionally, the assessment concluded that any potential Tave effect on a main steam line isolation was not a concern since the spikes were in the negative direction, thus towards the main steam line isolation setpoint.

The licensee's resolution of the inadvertent rod motion had not reached a final determination at the conclusion of this inspection.

C.

Conclusions The inspectors concluded that the licensee's operability assessment adequately addressed UFSAR accident analysis concerns regarding over power delta T and over temperature delta T reactor trip setpoints, and verified that the temperature spikes did not result from an instrument malfunction.

The licensee's actions to further evaluate frequent automatic rod motion for potential system failure were appropriate.

Operator Know(edge and Performance 04.1

~ 0 erator Res onse to a Service Water SW Pum Tri

Ins ection Sco e (71707)

The inspectors reviewed the performance of plant operators in response to an automatic trip of the C-SW pump.

b.

Observations and Findin s On December 25, 1996, control room operators received an alarm on annunciator J-9, "Safeguard Breaker Trip," and observed that the C-SW pump had tripped with its white light energized.

A screenhouse fire alarm (Z-26) also annunciated.

Operations personnel immediately started the D-SW pump to maintain flow in the affected train, and also dispatched the fire brigade.

The fire brigade reported no fire

in the screenhouse but did indicate that black smoke was observed in the area of the C-SW pump motor and that a smell of smoke was evident.

Operators placed the C-SW pump's control switch in "pull-to-stop," and its power supply circuit breaker was racked out.

As a result of this event, operations issued a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> event notification at approximately 1:17 p.m. under 10 CFR 50.72 (b)(2)(ii) (actuation of an engineered safety feature (ESF)).

Operations personnel indicated that the notification was issued based on the fact that the D-SW pump was selected to automatically actuate on an ESF signal, and was ESF equipment that was manually started.

Since no manual or automatic ESF actuation signal was actually generated, the licensee deemed that the notification was unnecessary, and it was later rescinded on January 2, 1997, As part of the licensee's ongoing effort to upgrade all four SW pumps with larger 350 horsepower motors, the C-SW motor had been scheduled to be replaced in early 1997.

However, after this failure, the licensee accelerated the motor replacement to December 26, 1996.

The failed motor was removed, dismantled, and inspected by RGSE maintenance personnel, with a representative from the Schultz Electric Corporation.

These individuals indicated that the failure appeared to be a phase-to-phase winding short.

The initial indications suggested that overheating of the windings may have occurred due to excess winding insulation applied when the motor was re-wound approximately one year ago.

The additional insulation was installed to reduce the magnitude of heat concentration in the motor windings, but instead appeared to reduce the rate of heat dissipation.

However, the specific root cause would be determined at a later time by a detailed inspection of the motor at the vendor's plant.

C.

Conclusions The inspectors concluded that control room operators and the fire brigade responded appropriately to the C-SW pump trip. Service water system flow was returned to normal immediately, and the cause of the screenhouse fire alarm was investigated and identified within five minutes of annunciation.

Additionally, the initial inspections of the failed pump motor indicated the primary cause as motor winding failure; a specific root cause analysis would be performed at the vendor's plant.

The installation of an upgraded motor should reduce the potential for recurrence.

Miscellaneous Operations Issues (92700)

08.1 0 en LER 96-013: Circuit Breakers Closed While in Mode 3 Due to Personnel Error Resulted in Condition Prohibited b Technical S ecifications.

LER 96-013 was submitted on November 27, 1996, in response to the closure of the breakers for safety injection (Sl) pump discharge valves prior to entry into Mode 4 during a reactor plant cooldown (see IR 50-244/96-11).

The breakers were shut for a short period of time with the plant about 3'F above the maximum average

RCS temperature (Tave) allowed for entry into Mode 4 ((350'F).

However, this resulted in two trains of the emergency core cooling system (ECCS) being considered inoperable, which required entry into ITS limiting condition for operation (LCO) 3.0.3.

The LER attributed the underlying cause of this event to personnel error, resulting from less than adequate verbal communications between the control room and in-plant operators, and the utilization of unclear supervisory methods.

The licensee initiated a Human Performance Enhancement System (HPES) evaluation in response to this event.

Pending completion of the evaluation, this LER remains open.

08.2 Closed LER 96-014: Pressure Relievin Ca abilit Could be De raded Due to Sin le Failure of DC Power Which Could Prevent Miti atin the Conse uences of an Accident.

LER 96-014 was submitted on December 5, 1996, in response to a discovery by the licensee's engineering department that loss of a single train of DC power could prevent the pressure relieving capability of the pressurizer power operated relief valves (PORVs) if one PORV block valve was already shut (see IR 50-244/96-011).

The PORVs were credited for depressurizing the primary plant in the event of a steam generator tube rupture (SGTR).

The LER attributed the cause of the event to an inadequate review of plant design features when the SGTR analysis was adopted in 1987.

The licensee notified the industry of this condition via the Nuclear NETWORK. The plant modification performed corrected the deficiency by providing the same DC power supply to a PORV and its associated block valve.

The inspectors considered that the LER adequately described this event and appropriately addressed the root causes and corrective actions.

This LER is closed.

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

The inspectors observed portions of plant maintenance activities to verify that the correct parts and tools were utilized, the applicable industry codes and technical specification requirements were satisfied, adequate measures were in place to ensure personnel safety and prevent damage to plant structures, systems, and components, and to ensure that equipment operability was verified upon completion of post maintenance testin b.

Observations and Findin s The inspectors observed portions of the following maintenance activities:

SW pump intermediate building control switch bypass modification, observed from December 2, 1996 - December 5, 1996 (see section M2.1)

B-emergency diesel generator output breaker troubleshooting observed from December 26, 1996 - January 4, 1997 (see section M2.2)

Troubleshooting and testing of HCV-484, steam dump pressure controller, observed from December 17, 1996 - December 20, 1996 (see section M2.3)

c.

Conclusions Overall, the inspectors concluded that the observed maintenance activities were performed effectively and were well coordinated with all site organizations.

Maintenance technicians demonstrated good technical abilities and performed their work.in accordance with procedure requirements.

M1.2 General Comments on Surveillance Activities a.

Ins ection Sco e (61726)

The inspectors observed selected surveillance tests to determine whether approved procedures were in use, details were adequate, test instrumentation was properly calibrated and used, technical specifications limiting conditions for operation (LCOs)

were satisfied, testing was performed by qualified personnel, test results satisfied acceptance criteria or were properly dispositioned, and correct post-test system restoration was performed.

b.

Observations and Findin s The inspectors observed portions of the following surveillance tests:

PT-27.4, "Diesel Generator Operation," observed on December 26, 1996.

This surveillance was performed to demonstrate the A-emergency diesel generator (EDG) operable following the failure of the B-EDG to pass its periodic test (PT-12.2) that same day.

The test verified that the A-EDG would start from the control room and that its associated feeder breakers to the plant's safeguard buses would shut.

No discrepancies were noted during the test.

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PT-16Q-T, "AuxiliaryFeedwater Turbine Pump-Quarterly," observed on January 2, 1997.

This quarterly surveillance was performed to verify operability of the turbine-driven auxiliary feedwater (TDAFW) pump.

The pump was tested at a total flow rate of 400 gallons per minute (gpm), and

the stroke times of the suction, discharge and steam supply valves were monitored and recorded.

No discrepancies were noted during the test.

c.

Conclusions Overall, the inspectors concluded that the observed surveillance activities were well controlled and coordinated with other site organizations.

Testing was performed in accordance with procedure requirements and technicians demonstrated a good understanding of the functional requirements of the equipment being tested.

M2 Maintenance and IVlaterial Condition of Facilities and Equipment M2.1 Failure of the B-SW Pum Breaker Durin Post-Maintenance Testin ar Ins ection Sco e (62707)

b.

The inspectors observed the licensee's activities in response to the B-SW pump post-modification test failure following removal of a control power transfer switch.

Observations and Findin s On December 12, 1996, the licensee completed a modification to remove the control power transfer switch for the B-SW pump (see section E2.1). The pump was then functionally tested in accordance with station modification procedure SM-96-121.2, "Service Water Pump 18 Remote Control Switch Removal."

The procedure included a step to verify the breaker's closing function by placing the breaker in the test position and having operations personnel close the breaker from the control room.

The breaker failed to close during this test and the licensee issued an ACTION Report (96-1218) to initiate a root cause analysis and corrective action.

The breaker was subsequently racked out and inspected by maintenance personnel.

Inspection revealed that one of the secondary electrical contacts had been deformed.

The contact was replaced and the breaker was re-tested on December 13, 1996.

The breaker operated as required and the B-SW pump was subsequently declared operable.

The inspectors were concerned that breaker contact deformity had occurred previously, specifically when troubleshooting a performance test failure for the B-EDG (see IR 50-244/96-11).

The inspectors were also concerned that this breaker design (Westinghouse DB-type) was also used with other safety related equipment in the plant.

The licensee indicated that a thorough root cause analysis would be performed as part of the ACTION Report process.

C.

Conclusions The inspectors concluded that maintenance personnel responded appropriately to the breaker failure, and successfully identified and corrected the deficiency in the breaker.

The inspectors determined that initiation of a thorough root cause analysis

to identify any existing common failure mechanisms for similar breakers in the plant was appropriate.

M2.2 B-Emer enc Diesel Generator B-EDG Out ut Breaker Troubleshootin

lns ection Sco e (62707)

The inspectors observed and reviewed corrective maintenance associated with a failure of the B-EDG output breaker to close.

b.

Observations and Findin s On December 26, 1996, the 8-EDG output breaker to safeguards bus 16 failed to shut in the test position upon demand from the control room during performance of PT-12.2, "Emergency Diesel Generator B." This breaker has had a recent history of failing to shut, and a significant amount of troubleshooting and corrective maintenance had previously been performed to address its performance (see IR 50-244/96-11).

The inspectors were concerned that this breaker has failed to pass its periodic test (PT) three times since October 28, 1996.

Following the failure on December 26, the breaker was visually inspected and verified for proper alignment.

Additional troubleshooting then concentrated on the control circuitry for the breaker.

The manual control switch was suspected as a possible cause for the PT failure based on continuity measurements taken during the troubleshooting efforts. The switch was removed, inspected, and bench tested by maintenance and engineering personnel.

These individuals concluded that replacement of the switch was necessary, based partly on the discovery of some corrosion on the contacts inside the switch, and partly due to an intermittent continuity failure identified during bench testing.

Failure of the control switch for the bus 16 breaker would not prevent the B-EDG from starting if an ESF actuation were to occur, as it is electrically out of the circuit during automatic actuation.

A new switch was installed on December 26, 1996.

PT-12.2 was then successfully performed, and the B-EDG was declared operable.

The licensee indicated that the PT would be re-performed with test equipment installed early in January 1996.

C.

Conclusions The inspectors concluded that maintenance and engineering personnel worked well together, and in a short time identified and resolved an equipment problem with the B-EDG. The licensee's plan to perform another PT with diagnostic test equipment was appropriate since the breaker has had recent test failure M2.3 Steam Dum Valve Pressure Controller Re lacement a.

Ins ection Sco e (62707)

The inspectors observed the troubleshooting, installation, and testing of steam dump valve pressure controller HCV-484.

b.

Observations and Findin s During the plant startup on November 12, 1996, steam dump valve controller HCV-484 was unable to maintain main steam system pressure in the "steam pressure mode."

The licensee concluded that the controller's setpoint was drifting and could not be adequately adjusted.

This required an additional operator to manually control the steam dump valves.

The licensee intended to replace the controller with an upgraded Westinghouse model (PID 9000).

However, the licensee indicated that it may take several months before the replacement controller could be delivered, and therefore decided to troubleshoot and repair the installed controller as an interim corrective action.

Troubleshooting revealed corroded contacts in the local/remote switch on the controller.

Additionally, during bench testing, the controller output was erratic and worsened when the local/remote switch was touched.

The switch was replaced and the controller was successfully bench tested, providing a steady output with zero deviation.

The controller was reinstalled on December 17, 1996, and final operability testing was conducted on December 20, 1996.

The instrumentation and control (IRC)

foreman conducted a detailed briefing with operations personnel to ensure everyone involved in the testing understood the test process, as operability would be determined by actual manipulation of HCV-484 in the "steam pressure mode."

The steam dump valves were not affected during the testing due to the valves being selected to automatic on the main control board.

The output of the controller was measured by installed test equipment, and the controller once again provided a steady output with zero deviation.

It was subsequently declared operable.

The licensee indicated that the Westinghouse controller would not be immediately installed upon delivery, based upon the successful troubleshooting and operability test results for the existing controller.

c.

Conclusions The inspectors concluded that ISC personnel successfully identified and resolved the operability problems with the steam dump valve controller.

Additionally, the briefing conducted between ISC and operations personnel was thorough and demonstrated good coordination and communication between the two department M2.4 B-Char in Pum Packin Leaka e and Char in S sterr. Pi in Vibrations aO Ins ection Sco e (62707)

The inspectors reviewed the licensee's response to increased packing leakage on the B-charging pump and to charging system piping vibrations.

b.

Observations and Findin s On December 12, 1996, operations personnel noted an increase in the charging pump packing leak rate from 0.075 to 0.210 gallons per minute (gpm) over the course of one 8-hour shift. The licensee attributed the source of this packing leakage to the B-charging pump, based on the different leak rates observed when the three charging pumps were run in different combinations.

The licensee maintains an administrative limitof 0.5 gpm total charging pump packing leakage when the charging system is in operation.

Charging pump packing leakage is considered RCS identified leakage and the leakage limit is based on the maximum amount of RCS identified leakage allowable (10 gpm) by the improved technical specifications (ITS). Even though charging pump system leakage was still below the administrative limit, the licensee decided to remove the B-charging pump from service to replace its packing.

On December 19, 1996, the B-charging pump was removed from service and its packing was replaced.

The pump was returned to service the following day, and total charging system leak rate was reduced to less than 0.1 gpm.

On December 13, 1996, while investigating the increased packing leakage, operations personnel noted that the suction piping on the C-charging pump was vibrating and initiated an ACTION Report (96-1226) to address the issue.

High vibration on the charging lines had been noted in the past.

The licensee had previously generated an ACTION Report (96-0822) to address the problem, which appeared to come and go intermittently. The A-charging pump and the C-charging pump relief valves had been replaced and aligned to minimize piping stress and this appeared to lessen the problem, but not eliminate it completely.

The licensee indicated that another possible cause of the vibration was leakage through the B-charging pump discharge check valves, and also indicated that they intended to replace the valves during the next outage.

C.

Conclusions The inspectors concluded that the licensee took appropriate action in response to the noted increase in the charging pump leak rate.

The licensee's corrective actions were considered successful in that the measured leak rate dropped significantly following the B-charging p>>mp packing replacement.

The inspectors also concluded that the licensee's efforts to reduce charging system piping vibrations had reduced the magnitude of the problem, but that the root cause of the problem had not yet been fully identifie III. En ineerin E2 Engineering Support of Facilities and Equipment E2.1 Service Water Pum Control Power Transfer Switch Modification a.

Ins ection Sco e (37551)

The inspectors reviewed the licensee's engineering modification to remove the SW pump control power transfer switches in the intermediate building.

b.

Observations and Findin s On November 25, 1996, the licensee generated an ACTION Report (96-1125) to address the possibility that the SW control power transfer switches, and local wiring and start buttons, could fail as a result of the intermediate building block wall collapsing in the event of a high energy line break.

The licensee concluded that a block wall failure could disable all DC control power to the SW pumps and prevent automatic start of the pumps, which are necessary for cooling the emergency diesel generators in the event of a coincident loss of all offsite power.

The licensee performed a safety evaluation and determined that the SW pump control power transfer switches and push buttons were identified in the Updated Final Safety Analysis Report (UFSAR) as being for convenience only, and that these devices were not needed to respond to any plant transient or accident condition.

Additionally, the switches were not credited under any 10 CFR 50, Appendix R scenario, where starting the SW pumps would be accomplished by locally pulling the DC control fuses and closing the power supply breakers in the screenhouse.

The licensee therefore decided to bypass the switches by splicing the control power cables in the cable tunnel.

Work commenced on December 5, 1996, and was completed on December 16, 1996.

One SW pump breaker failed to operate during the post modification testing; however, this test failure was unrelated to the modification (see section M2.1).

c.

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Conclusions The inspectors concluded that the identification of this problem was a good finding by the licensee's engineering department.

The inspectors also concluded that the modification adequately resolved the concern that a potential loss of all service water could occur should the intermediate building block wall fail. The licensee's safety evaluation was appropriately focused on accident and fire protection issues.

E2,2 Potential Over ressurization of the Containment Charcoal Filter Dousin Header a.

Ins ection Sco e (37551)

The inspectors reviewed the licensee's resolution to a potential overpressurization of the containment charcoal filter dousing header during a design base acciden b.

Observations and Findin s On December 20, 1996, while reviewing issues pertaining to Generic Letter (GL) 96-06, "Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions," the licensee identified that the possibility existed for a thermally induced overpressurization that could fail the charcoal filter dousing lines between the check valves in the cross-connect piping for the A-and B-containment spray (CS) headers.

This failure could occur from a heatup of the water-filled line following a design basis loss of coolant accident inside containment.

Such an overpressurization could render both trains of containment spray incapable of performing their safety function because of the excessive flow diversion from the CS system through the break.

The dousing function was not credited in the Ginna Accident Analysis, and the licensee therefore isolated and vented the line on December 21, 1996.

On December 23, 1996, the licensee completed calculations confirming that a thermally induced overpressurization in the dousing line was possible, and reported the condition to the NRC under 10 CFR 50.72 (b)(2)(iii)(D) as a condition that could prevent a safety function from mitigating the consequences of an accident.

On January 3, 1997, the licensee installed a thermal relief valve in the containment spray charcoal filter dousing lines to prevent a thermal pressurization that would overstress the dousing lines.

The relief valve was set for a maximum relief capacity of 10 gpm at 500 psig to prevent pressure from exceeding the rated maximum piping design pressure.

After the relief valve was installed, the dousing line was unisolated.

C.

Conclusions The inspectors concluded that the licensee's efforts in identifying and evaluating the potential for overpressurization of the containment charcoal filter dousing header were both effective and timely. The final resolution to install a relief valve in the line was appropriate, in that the safety function of the containment spray system could be maintained without eliminating the capability to douse the charcoal filters.

E2.3, Criticalit Monitor For The New Fuel Pre aration Area aO Ins ection Sco e (37551)

The inspector reviewed the licensee's ongoing efforts to evaluate the use of an installed area radiation monitor to meet the requirements of 10 CFR 70.24,

"CriticalityAccident Requirements."

b.

Observations and Findin s NRC Inspection Report 50-244/96-07 reported the licensee's efforts to evaluate use of the area radiation monitor (R-5) installed near the spent fuel pool in the auxiliary building as a criticality monitor for the new fuel preparation area (NFPA). There is currently no instrument installed in the auxiliary building that can meet the specific

requirements contained in 10 CFR 70.24 for monitoring an inadvertent criticality in the NFPA. The rule requires that facilities licensed to possess greater than 700 grams of uranium-235 have the ability to detect specific radiation levels at defined distances from a critical source.

During the NRC Systematic Evaluation Program (SEP) reviews of the Ginna plant in the early 1980s, the NRC questioned the licensee's ability to monitor for a criticality in the NFPA. The licensee indicated that the NFPA had an inherently safe design that would ensure subcritical conditions in stored new fuel at all times.

The geometric arrangement of the storage racks and their seismic design prevent the assembly of a critical mass, and the location and arrangement of the NFPA enclosure provides for protection against flooding.

As a result of these design features, a criticality monitor was not considered necessary, and the NRC accepted the licensee's response.

Although the R-5 instrument is set at 25 mR/hr, and would alarm from a criticality in the NFPA, the licensee agreed in August 1996 to evaluate the instrument for meeting the detailed specifications of 10 CFR 70.24.

Evaluations conducted by RGSE engineering and radiological protection personnel are inconclusive to date, but have indicated that the R-5 monitor may be reset from it's present alarm setpoint to meet the specifications of the rule. The evaluation is still ongoing because the configuration of the NFPA is such that the radiation field from a criticality could vary significantly depending on the exact location of the source material, and this could present a difficultlyfor determining the proper alarm set point. Also, some portions of the NFPA are shielded from R-5, which could complicate the ability of the instrument to detect a criticality under the rule's specification.

The licensee anticipated that the evaluation will also determine if installation of additional monitors near the NPFA is feasible, or if there is a justifiable cause for seeking an exemption, as allowed by the rule.

No new fuel will be stored in the auxiliary building before July 1997.

Conclusions The inspector concluded that the licensee's ongoing efforts were well focused on meeting the specifications of 10 CFR 70.24.

However, it is not apparent that strict conformance to the rule is required since the NRC had previously accepted the existing monitor.

Pending the outcome of the licensee's evaluation and additional review by the NRC, this item is unresolved (URI 50-244/96-12-01).

IV. Plant Su ort P1 Radiological Protection and Chemistry (RPSC) Controls R1.1 Contamination Bounda Control and Radiolo ical Labelin a.

Ins ection Sco e (71750)

The inspectors reviewed the licensee's follow-up actions in response to contamination boundary control problems and radiological labeling deficiencies identified by the NRC in IR 50-244/96-11.

b.

Observations and Findin s On October 28, 1996, during a tour of the auxiliary building, NRC inspectors discovered some improper use of boundary controls in a contaminated area.

The inspectors identified several examples of inadequate contamination boundary controls and several radiological labelling problems throughout the auxiliary building.

The licensee generated two ACTION Reports (96-0993 & 96-0983) to address the conditions.

From November 1, 1996, to November 15, 1996, the licensee upgraded all old or faded radiological labels in the auxiliary building with new ones, and reduced the total number of labels to the minimum required to properly identify auxiliary building contamination area boundary. postings and radioactive material containers.

The licensee concluded that the contamination boundary control problems occurred as a result of poor radiological work practices in combination with poor management oversight of the work area.

The licensee indicated that the radiation protection staff would conduct regular tours to monitor effective contamination area boundary control.

In addition, the licensee indicated that they intend to incorporate enhanced training on contamination boundary control into their "advanced radiation worker" as well as their maintenance

"toolbox" training.

The inspectors questioned whether contamination boundary control problems could recur in the future, since no formal programmatic upgrades to the existing boundary controls had been made.

However, the licensee considered that the identified corrective actions were sufficient to address these problems.

C.

Conclusions The inspectors concluded that the licensee successfully corrected the inadequate boundary controls and resolved the radiological labelling deficiencies in the auxiliary building.

Increased supervisory surveillance of work inside contamination areas should help maintain proper contamination boundary controls.

However, the inspectors noted these changes were not made as formal programmatic upgrades, and the inspectors questioned whether they would prevent future recurrence of contamination boundary control problem V. IVlana ement Meetin s X1 Exit Me~ting Summary The inspectors presented the inspection results to members of licensee management on January 10, 1997.

The licensee acknowledged the findings presented, and indicated that none of the materials examined during the inspection were considered proprietary.

L1 Review of UFSAR Commitments A recent discovery of a licensee operating its facility in a manner contrary to the Updated Final Safety Analysis Report (UFSARI description highlighted the need for a special focused review that compares plant practices, procedures and/or parameters to the UFSAR description.

While performing the inspections discussed in this report, the inspectors reviewed the applicable portions of the UFSAR that related to the areas inspected to verify that the UFSAR wording was consistent with the observed plant practices, procedures and/or parameters.

No discrepancies were noted;

Attachment

PARTIAL LIST OF PERSONS CONTACTED Licensee P. Bamford B. Flynn A. Harhay J. Hotchkiss R. Marchionda W. Rapin J. Smith J. Widay T. White G. Wrobel Systems Engineer Primary Systems Engineering Manager Chemistry 5 Radiological Protection Manager Mechanical Maintenance Manager Production Superintendent Systems Engineer Maintenance Superintendent Plant Manager Operations Manager Nuclear Safety 5 Licensing Manager INSPECTION PROCEDURES USED IP 37551:

IP 62707:

IP 61726:

IP 71750:

IP 71707:

IP 83750:

IP 92700:

IP 92902:

IP 92903:

Onsite Engineering

'aintenance Observation Surveillance Observation Plant Support Activities Plant Operations Occupational Exposure Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities Follow-up - Engineering Follow-up - Maintenance ITEMS OPENED AND CLOSED

~Oen LER 96-013 Circuit Breakers Closed While in Mode 3, Due to Personnel Error, Resulted in Condition Prohibited by Technical Specifications.

URI 50-244/96-1 2-01 Requirements of 10 CFR 70.24, "CriticalityAccident Requirements."

Closed LER 96-014 Pressure Relieving Capability Could be Degraded Due to Single Failure of DC Power, Which Could Prevent Mitigating the Consequences of an Acciden Attachment

LIST OF ACRONYMS

~Acron m Definition AFW AOV CFR ECCS EDG EP ESF Fl FIC GL gpm HCV HPES I&C IEEE IP IR ITS LCO LER MDAFW MOV MOVATS NFPA mR/hr NS&L PORC PORV PT Pl PWR RCS RG&E RP&C RTD RWST SEP SEV SGTR Sl STA SW T

TSR Auxiliary Feedwater Air-Operated Valve Code of Federal Regulations Emergency Core Cooling System Emergency Diesel Generator Emergency Preparedness Engineered Safety Feature Flow Indicator Flow Indicator Controller Generic Letter Gallons per Minute Hydraulic Control Valve Human Performance Evaluation System Instrumentation and Control Institute of Electrical and Electronics Engineers Inspection Procedure

Inspection Report

Improved Technical Specifications

Limiting Condition for Operation

Licensee Event Report

Motor-Driven Auxiliary Feedwater

Motor-Operated Valve

MOV Analysis and Test System

New Fuel Preparation Area

milli Rem per hour

Nuclear Safety & Licensing

Plant Operations Review Committee

Power-Operated

Relief Valve

Periodic Test

Pressure

Indicator

Pressurized Water Reactor

Reactor Coolant System

Rochester

Gas and Electric Corporation

Radiological Protection and Control

Resistance

Temperature

Detector

Refueling Water Storage Tank

Systematic Evaluation Program

Safety Evaluation

Steam Generator Tube Rupture

Safety Injection

Shift Technical Advisor

Service Water

Temperature

Technical Services Request

'(

Attachment

TDAPN

l'cgAR

Turbine-Driven Auxiliary Feedwater

Updated Final Safety Analysis Report

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